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Sample records for axial burnup profile

  1. 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.

  2. 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.

  3. BWR AXIAL PROFILE

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    J. Huffer

    2004-09-28

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I.

  4. 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

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

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    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.

  6. 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

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    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)

  7. 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)

  8. 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)

  9. 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.

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

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    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)

  11. Axial Compressive Strength of Foamcrete with Different Profiles and Dimensions

    Directory of Open Access Journals (Sweden)

    Othuman Mydin M.A.

    2014-01-01

    Full Text Available Lightweight foamcrete is a versatile material; primarily consist of a cement based mortar mixed with at least 20% volume of air. High flow ability, lower self-weight, minimal requirement of aggregate, controlled low strength and good thermal insulation properties are a few characteristics of foamcrete. Its dry densities, typically, is below 1600kg/m3 with compressive strengths maximum of 15MPa. The ASTM standard provision specifies a correction factor for concrete strengths of between 14 and 42MPa to compensate for the reduced strength when the aspect height-to-diameter ratio of specimen is less than 2.0, while the CEB-FIP provision specifically mentions the ratio of 150 x 300mm cylinder strength to 150 mm cube strength. However, both provisions requirements do not specifically clarify the applicability and/or modification of the correction factors for the compressive strength of foamcrete. This proposed laboratory work is intended to study the effect of different dimensions and profiles on the axial compressive strength of concrete. Specimens of various dimensions and profiles are cast with square and circular cross-sections i.e., cubes, prisms and cylinders, and to investigate their behavior in compression strength at 7 and 28 days. Hypothetically, compressive strength will decrease with the increase of concrete specimen dimension and concrete specimen with cube profile would yield comparable compressive strength to cylinder (100 x 100 x 100mm cube to 100dia x 200mm cylinder.

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

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    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

  13. Experimental - theoretical study of axially compressed cold formed steel profiles

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    Bešević Miroslav

    2011-01-01

    Full Text Available Analysis of axially compressed steel members made of cold formed profiles presented in this paper was conducted through both experimental and numerical methods. Numerical analysis was conducted by means of "PAK" finite element software designed for nonlinear static and dynamic analysis of structures. Results of numerical analysis included ultimate bearing capacity with corresponding middle section force-deflection graphs and buckling curves. Extensive experimental investigation were also concentrated on determination of bearing capacity and buckling curves. Experiments were conducted on five series with six specimens each for slenderness values of 50, 70, 90, 110 and 120. Compressed simply supported members were analyzed on Amsler Spherical pin support with unique electronical equipment and software. Besides determination of forcedeflection curves, strains were measured in 18 or 12 cross sections along the height of the members. Analysis included comparisons with results obtained by different authors in this field recently published in international journals. Special attention was dedicated to experiments conducted on high strength and stainless steel members.

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

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    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)

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

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    Enercon Services, Inc.

    2011-03-14

    compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the costs of code data library updating, the use of S/U analysis methodologies could be accomplished on a shorter schedule and a lower cost than the gathering of sufficient experimental data. ENERCON estimates of the costs of an updated S/U computer code and data suite are $5M to $10M with a schedule of two to three years. Recent ORNL analyses using the S/U analysis method show that the bias and uncertainty values for fission product cross sections are smaller than previously expected. This result is confirmed by a similar EPRI approach using different data and computer codes. ENERCON also found that some issues regarding the implementation of burnup credit appear to have been successfully resolved especially the axial burnup profile issue and the depletion parameter issue. These issues were resolved through data gathering activities at the Yucca Mountain Project and ORNL.

  16. The Sensitivity Analysis of Axial Pressure Tube Creep Profile for Dryout Power in PHWR

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    Ryu, Euiseung; Kim, Youngae [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Stern Laboratory performed the CHF tests with only one axial pressure tube creep profile per 3.3%, 5.1% peak crept channel and made CHF correlation including creep factor from the CHF test results. Wolsong nuclear power plants also have utilized the same CHF correlation derived by CNL. Pressure tube diameter creep rate is function of fast neutron, coolant temperature, and coolant pressure in a channel. It means that various axial pressure tube creep profiles exist in PHWR due to the history of operating conditions. Usually, CHF correlation is used during ROP(Regional Overpower Protection) Trip Setpoint Analysis or Safety Analysis in PHWR. The sensitivity analysis for CHF effects using various creep profiles is needed. This paper summarizes the comparison results of dryout power between CHF test creep profile and estimated creep profiles of Wolsong units. The effect of axial pressure tube creep profile for dryout power in fuel channel is evaluated by using Stern Lab. CHF test creep profile and 380 channel creep profiles of Wolsong. The dryout powers at 3.3% and 5.1% test conditions are slightly smaller when using 380 Wolsong channels creep profiles. These also show that the simulated dryout powers maintain consistency regardless of flow conditions.

  17. Low-pressure reversible axial fan with straight profile blades and relatively high efficiency

    OpenAIRE

    2012-01-01

    The paper presents the design and operating characteristics of a model of reversible axial fan with only one impeller, whose reversibility is achieved by changing the direction of rotation. The fan is designed for the purpose of providing alternating air circulation in wood dryers in order to reduce the consumption of electricity for the fan and increase energy efficiency of the entire dryer. To satisfy the reversibility of flow, the shape of the blade profile is symmetrical along the l...

  18. Low-pressure reversible axial fan with straight profile blades and relatively high efficiency

    OpenAIRE

    Spasić Živan T.; Milanović Saša M.; Šušteršič Vanja M.; Nikolić Boban D.

    2012-01-01

    The paper presents the design and operating characteristics of a model of reversible axial fan with only one impeller, whose reversibility is achieved by changing the direction of rotation. The fan is designed for the purpose of providing alternating air circulation in wood dryers in order to reduce the consumption of electricity for the fan and increase energy efficiency of the entire dryer. To satisfy the reversibility of flow, the shape of the blade profile is symmetrical along the l...

  19. Low-pressure reversible axial fan with straight profile blades and relatively high efficiency

    Directory of Open Access Journals (Sweden)

    Spasić Živan T.

    2012-01-01

    Full Text Available The paper presents the design and operating characteristics of a model of reversible axial fan with only one impeller, whose reversibility is achieved by changing the direction of rotation. The fan is designed for the purpose of providing alternating air circulation in wood dryers in order to reduce the consumption of electricity for the fan and increase energy efficiency of the entire dryer. To satisfy the reversibility of flow, the shape of the blade profile is symmetrical along the longitudinal and transversal axes of the profile. The fan is designed with equal specific work of all elementary stages, using the method of lift forces. The impeller blades have straight mean line profiles. The shape of the blade profile was adopted after the numerical simulations were carried out and high efficiency was achieved. Based on the calculation and conducted numerical simulations, a physical model of the fan was created and tested on a standard test rig, with air loading at the suction side of the fan. The operating characteristics are shown for different blade angles. The obtained maximum efficiency was around 0.65, which represents a rather high value for axial fans with straight profile blades.

  20. Dose profile measurement in computerized axial tomography equipment using thermoluminescent dosemeters; Medicion del perfil de dosis en equipos de tomografia axial computarizada usando dosimetros termoluminiscentes

    Energy Technology Data Exchange (ETDEWEB)

    Azorin V, J.C.; Falcony, C.; Azorin N, J. [Centro de Investigacion en Ciencia y Tecnologia Avanzada, IPN, 07000 Mexico D.F. (Mexico)

    2000-07-01

    In this work are presented the results about measuring the radiation dose profile in two equipment of computerized axial tomography (Tac). Thermoluminescent dosemeters (Dtl) of LiF, Mg, Cu, P + Ptfe in form of disks were used which were developed and made in Mexico. The results showed that Dtl are appropriated for these type of studies. (Author)

  1. Operating characteristics of heavy loaded cylindrical journal bearing with variable axial profile

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    Stanislaw Strzelecki

    2005-12-01

    Full Text Available During the operation of turbounit its bearings displace as a result of heat elongation of bearings supports. It changes the static deflection line of rotor determined during assembly of the turbounit, causing an increase in the stresses on the bearing edges and a decrease in the dynamic state of the machine. One of possibilities to avoid the edge stresses is to apply the bearings with variable axial profile, e.g. hyperboloidal, convex profile in the axial cross-section of bearing. Application of journal bearings with hyperboloidal profile allows to extend the bearing operation range without the stress concentration on the edges of bush. These bearings successfully carry the extreme load in conditions of misaligned axis of journal and the bush eliminating the necessity of using self-aligning bearings. Operating characteristics of bearing include the resulting force, attitude angle, oil film pressure and temperature distributions, minimum oil film thickness, maximum oil film temperature. In literature there is a lack of data on the operating characteristics of heavy loaded hyperboloidal journal bearings operating at the conditions of adiabatic oil film and static equilibrium position of the journal. For the hyperboloidal bearing the operating characteristics have been obtained. Different values of length to diameter ratio, assumed shape and inclination ratio coefficients have been assumed. Iterative solution of the Reynolds', energy and viscosity equations was applied. Adiabatic oil film, laminar flow in the bearing gap as well as aligned and misaligned orientation of journal in the bush were considered.

  2. 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.

  3. 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)

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

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    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

  5. Fibromyalgia in patients with axial spondyloarthritis: epidemiological profile and effect on measures of disease activity.

    Science.gov (United States)

    Salaffi, Fausto; De Angelis, Rossella; Carotti, Marina; Gutierrez, Marwin; Sarzi-Puttini, Piercarlo; Atzeni, Fabiola

    2014-08-01

    To determine the prevalence of fibromyalgia (FM) in patients with ankylosing spondylitis (AS) or psoriatic arthritis (PsA) characterized by axial involvement (axial-PsA), and to assess the discriminative ability of different versions of the Ankylosing Spondylitis Disease Activity Score (ASDAS) and the Bath Ankylosing Spondylitis Activity Disease Activity Index (BASDAI) in measuring disease activity in three different cohorts of patients with axial spondyloarthritis (axial-SpA), FM, or both (axial-SpA + FM), this study was divided into two phases: (1) 402 patients with definite AS or axial-PsA were examined to diagnose FM and estimate its prevalence; and (2) 419 patients (111 with axial-SpA, 248 with FM, and 60 with aSpA + FM) were evaluated using the different versions of the ASDAS and BASDAI to assess the effect on disease activity. The overall prevalence of FM in the axial-SpA population was 14.9 %, significantly higher among women (p axial-PsA was 17.2 %. Although the BASDAI scores correlated with those of ASDAS-C-reactive protein (CRP) and ASDAS-erythrocyte sedimentation rate (ESR) (p axial-SpA and more prevalent in female patients. Our findings suggest that ASDAS is better than BASDAI in distinguishing patients with disease activity from those with functional impairment. The use of ASDAS may be very useful in clinical practice as it allows treating patients with the most appropriate therapy.

  6. Velocity profile of water vapor inside a cavity with two axial inlets and two outlets

    Science.gov (United States)

    Guadarrama-Cetina, José; Ruiz Chavarría, Gerardo

    2014-03-01

    To study the dynamics of Breath Figure phenomenon, a control of both the rate of flow and temperature of water vapor is required. The experimental setup widely used is a non hermetically closed chamber with cylindrical geometry and axial inlets and outlets. In this work we present measurements in a cylindrical chamber with diameter 10 cm and 1.5 cm height, keeping a constant temperature (10 °C). We are focused in the velocity field when a gradient of the temperatures is produced between the base plate and the vapor. With a flux of water vapor of 250 mil/min at room temperature (21 °C), the Reynolds number measured in one inlet is 755. Otherwise, the temperatures of water vapor varies from 21 to 40 °C. The velocity profile is obtained by hot wire anemometry. We identify the stagnations and the possibly instabilities regions for an empty plate and with a well defined shape obstacle as a fashion sample. Facultad de Ciencias, UNAM.

  7. Estimation of volume flow in curved tubes based on analytical and computational analysis of axial velocity profiles

    Science.gov (United States)

    Verkaik, A. C.; Beulen, B. W. A. M. M.; Bogaerds, A. C. B.; Rutten, M. C. M.; van de Vosse, F. N.

    2009-02-01

    To monitor biomechanical parameters related to cardiovascular disease, it is necessary to perform correct volume flow estimations of blood flow in arteries based on local blood velocity measurements. In clinical practice, estimates of flow are currently made using a straight-tube assumption, which may lead to inaccuracies since most arteries are curved. Therefore, this study will focus on the effect of curvature on the axial velocity profile for flow in a curved tube in order to find a new volume flow estimation method. The study is restricted to steady flow, enabling the use of analytical methods. First, analytical approximation methods for steady flow in curved tubes at low Dean numbers (Dn) and low curvature ratios (δ) are investigated. From the results a novel volume flow estimation method, the cos θ-method, is derived. Simulations for curved tube flow in the physiological range (1≤Dn≤1000 and 0.01≤δ≤0.16) are performed with a computational fluid dynamics (CFD) model. The asymmetric axial velocity profiles of the analytical approximation methods are compared with the velocity profiles of the CFD model. Next, the cos θ-method is validated and compared with the currently used Poiseuille method by using the CFD results as input. Comparison of the axial velocity profiles of the CFD model with the approximations derived by Topakoglu [J. Math. Mech. 16, 1321 (1967)] and Siggers and Waters [Phys. Fluids 17, 077102 (2005)] shows that the derived velocity profiles agree very well for Dn≤50 and are fair for 50100), no analytical approximation method exists. In the position of the maximum axial velocity, a shift toward the inside of the curve is observed for low Dean numbers, while for high Dean numbers, the position of the maximum velocity is located at the outer curve. When the position of the maximum velocity of the axial velocity profile is given as a function of the Reynolds number, a "zero-shift point" is found at Re=21.3. At this point the shift in

  8. 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)

  9. Tangential Velocity Profile for Axial Flow Through Two Concentric Rotating Cylinders with Radial Magnetic Field

    Directory of Open Access Journals (Sweden)

    Girishwar Nath

    1970-10-01

    Full Text Available A closed form solution of the Navier-Stokes equations has been obtained in the case of steady axisymmetric flow of an incompressible electrically conducting viscous fluid between two concentric rotating cylinders composed of an insulating material under the influence of radial magnetic field. It has been found that the velocity components are less than those of the classical hydrodynamic case. In the presence of the magnetic field, the tangential velocity becomes fully developed in a smaller axial distance than in the absence of the magnetic field. For small Reynolds number, the fully developed tangential velocity is achieved in a small axial distance, but it requires greater axial distance for large Reynolds number.

  10. Measurement of axial neutral density profiles in a microwave discharge ion thruster by laser absorption spectroscopy with optical fiber probes.

    Science.gov (United States)

    Tsukizaki, Ryudo; Koizumi, Hiroyuki; Nishiyama, Kazutaka; Kuninaka, Hitoshi

    2011-12-01

    In order to reveal the physical processes taking place within the "μ10" microwave discharge ion thruster, internal plasma diagnosis is indispensable. However, the ability of metallic probes to access microwave plasmas biased at a high voltage is limited from the standpoints of the disturbance created in the electric field and electrical isolation. In this study, the axial density profiles of excited neutral xenon were successfully measured under ion beam acceleration by using a novel laser absorption spectroscopy system. The target of the measurement was metastable Xe I 5p(5)((2)P(0) (3/2))6s[3/2](0) (2) which absorbed a wavelength of 823.16 nm. Signals from laser absorption spectroscopy that swept a single-mode optical fiber probe along the line of sight were differentiated and converted into axial number densities of the metastable neutral particles in the plasma source. These measurements revealed a 10(18) m(-3) order of metastable neutral particles situated in the waveguide, which caused two different modes during the operation of the μ10 thruster. This paper reports a novel spectroscopic measurement system with axial resolution for microwave plasma sources utilizing optical fiber probes.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. Azimuthal velocity profiles in Rayleigh-stable Taylor-Couette flow and implied axial angular momentum transport

    CERN Document Server

    Nordsiek, Freja; van der Veen, Roeland C A; Sun, Chao; Lohse, Detlef; Lathrop, Daniel P

    2014-01-01

    Azimuthal velocity profiles were measured in a Taylor-Couette apparatus, which has been used as a model of stellar and planetary accretion disks. The apparatus has a cylinder radius ratio of $\\eta = 0.7158$, an aspect-ratio of $\\Gamma = 11.74$, and axial boundaries attached to the outer cylinder --- known to have significant Ekman pumping. We investigated angular momentum transport and Ekman pumping in the Rayleigh-stable regime. The regime is linearly stable and is characterized by radially increasing specific angular momentum. We measured several Rayleigh-stable profiles for shear Reynolds numbers $Re_S \\sim O\\left(10^5\\right) \\,$, both for $\\Omega_i > \\Omega_o > 0$ (quasi-Keplerian regime) and $\\Omega_o > \\Omega_i > 0$ (sub-rotating regime) where $\\Omega_{i,o}$ is the inner/outer cylinder rotation rate. None of the velocity profiles matched the non-vortical laminar Taylor-Couette profile. The deviation from that profile increased as solid-body rotation was approached at fixed $Re_S$. Flow super-rotation, a...

  16. The influence of the tangential velocity of inner rotating wall on axial velocity profile of flow through vertical annular pipe with rotating inner surface

    Directory of Open Access Journals (Sweden)

    Sharf Abdusalam M.

    2014-03-01

    Full Text Available In the oil and gas industries, understanding the behaviour of a flow through an annulus gap in a vertical position, whose outer wall is stationary whilst the inner wall rotates, is a significantly important issue in drilling wells. The main emphasis is placed on experimental (using an available rig and computational (employing CFD software investigations into the effects of the rotation speed of the inner pipe on the axial velocity profiles. The measured axial velocity profiles, in the cases of low axial flow, show that the axial velocity is influenced by the rotation speed of the inner pipe in the region of almost 33% of the annulus near the inner pipe, and influenced inversely in the rest of the annulus. The position of the maximum axial velocity is shifted from the centre to be nearer the inner pipe, by increasing the rotation speed. However, in the case of higher flow, as the rotation speed increases, the axial velocity is reduced and the position of the maximum axial velocity is skewed towards the centre of the annulus. There is a reduction of the swirl velocity corresponding to the rise of the volumetric flow rate.

  17. Laser induced fluorescence measurements of axial velocity, velocity shear, and parallel ion temperature profiles during the route to plasma turbulence in a linear magnetized plasma device

    Science.gov (United States)

    Chakraborty Thakur, S.; Adriany, K.; Gosselin, J. J.; McKee, J.; Scime, E. E.; Sears, S. H.; Tynan, G. R.

    2016-11-01

    We report experimental measurements of the axial plasma flow and the parallel ion temperature in a magnetized linear plasma device. We used laser induced fluorescence to measure Doppler resolved ion velocity distribution functions in argon plasma to obtain spatially resolved axial velocities and parallel ion temperatures. We also show changes in the parallel velocity profiles during the transition from resistive drift wave dominated plasma to a state of weak turbulence driven by multiple plasma instabilities.

  18. 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.

  19. 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

  20. CLINICAL PROFILE OF PATIENTS WITH EARLY AXIAL SPONDYLOARTHRITIS (RUSSIAN COHORT OF PATIENTS

    Directory of Open Access Journals (Sweden)

    E. E. Gubar'

    2014-01-01

    Full Text Available Objective: to study clinical manifestations of axial spondyloarthritis (axSpA fulfilling ASAS criteria and to evaluate Russian version of modified New York criteria for the diagnosis of AS in Russian patients.Subjects and methods. Authors examined 73 patients aged 18–45 years suffering from inflammatory back pain for a period from 3 months to 5 years. BASDAI and ASDAS-CRP were used to assess activity, whereas BASFI – to evaluate functional status. Examination included: assessment of HLA-B27 rate, X-ray of pelvis and lumbar spine, ultrasonography of hip joints and calcaneal regions, magnetic-resonance imaging (MRI of sacroiliac joints, lumbar spine and hip joints (if clinical signs of injury are present, densitometry of lumbar spine (LII–IV and femoral neck.Results. Mean age of patients was 28.3±6.4 years, mean duration of disease – 19.9±14.4 months. HLA-B27 was found in 94.5% of patients. Mean BASDAI value was 4.1±1.9; ASDAS – 2.7±1.3; BASFI – 2.6±2.1. Peripheral arthritis was observed in 65.8% of cases, coxitis – in 31.5%, calcaneal enthesitis – in 61.6%, dactylitis – in 19.2%, low bone mineral density – in 17.8%. MRI showed inflammatory changes of axial skeleton in 84.9% of patients, active sacroiliitis (SI – in 72.6%. X-ray revealed definite SI in 49.3% of patients («classic» AS. According to MRI data, 30.1% of patients with active SI and without structural changes of sacroiliac joints had pre-radiological stage of AS (by Russian version of modified New York criteria. 74.0% of patients fulfilled both sets of ASAS criteria for axSpA, 5.5% – met only I criteria set, whereas 20.5% – only II criteria set. Three groups of patients were defined. The first included patients with radiologically proven SI, the second – with MRI-proven SI and the third – patients without SI. Significant difference between the groups was detected either by gender (number of males in groups I and II exceeded that in group

  1. In-shoe plantar tri-axial stress profiles during maximum-effort cutting maneuvers.

    Science.gov (United States)

    Cong, Yan; Lam, Wing Kai; Cheung, Jason Tak-Man; Zhang, Ming

    2014-12-18

    Soft tissue injuries, such as anterior cruciate ligament rupture, ankle sprain and foot skin problems, frequently occur during cutting maneuvers. These injuries are often regarded as associated with abnormal joint torque and interfacial friction caused by excessive external and in-shoe shear forces. This study simultaneously investigated the dynamic in-shoe localized plantar pressure and shear stress during lateral shuffling and 45° sidestep cutting maneuvers. Tri-axial force transducers were affixed at the first and second metatarsal heads, lateral forefoot, and heel regions in the midsole of a basketball shoe. Seventeen basketball players executed both cutting maneuvers with maximum efforts. Lateral shuffling cutting had a larger mediolateral braking force than 45° sidestep cutting. This large braking force was concentrated at the first metatarsal head, as indicated by its maximum medial shear stress (312.2 ± 157.0 kPa). During propulsion phase, peak shear stress occurred at the second metatarsal head (271.3 ± 124.3 kPa). Compared with lateral shuffling cutting, 45° sidestep cutting produced larger peak propulsion shear stress (463.0 ± 272.6 kPa) but smaller peak braking shear stress (184.8 ± 181.7 kPa), of which both were found at the first metatarsal head. During both cutting maneuvers, maximum medial and posterior shear stress occurred at the first metatarsal head, whereas maximum pressure occurred at the second metatarsal head. The first and second metatarsal heads sustained relatively high pressure and shear stress and were expected to be susceptible to plantar tissue discomfort or injury. Due to different stress distribution, distinct pressure and shear cushioning mechanisms in basketball footwear might be considered over different foot regions.

  2. 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.

  3. Numerical performance analysis of acoustic Doppler velocity profilers in the wake of an axial-flow marine hydrokinetic turbine

    Energy Technology Data Exchange (ETDEWEB)

    Richmond, Marshall C.; Harding, Samuel F.; Romero Gomez, Pedro DJ

    2015-09-01

    The use of acoustic Doppler current profilers (ADCPs) for the characterization of flow conditions in the vicinity of both experimental and full scale marine hydrokinetic (MHK) turbines is becoming increasingly prevalent. The computation of a three dimensional velocity measurement from divergent acoustic beams requires the assumption that the flow conditions are homogeneous between all beams at a particular axial distance from the instrument. In the near wake of MHK devices, the mean fluid motion is observed to be highly spatially dependent as a result of torque generation and energy extraction. This paper examines the performance of ADCP measurements in such scenarios through the modelling of a virtual ADCP (VADCP) instrument in the velocity field in the wake of an MHK turbine resolved using unsteady computational fluid dynamics (CFD). This is achieved by sampling the CFD velocity field at equivalent locations to the sample bins of an ADCP and performing the coordinate transformation from beam coordinates to instrument coordinates and finally to global coordinates. The error in the mean velocity calculated by the VADCP relative to the reference velocity along the instrument axis is calculated for a range of instrument locations and orientations. The stream-wise velocity deficit and tangential swirl velocity caused by the rotor rotation lead to significant misrepresentation of the true flow velocity profiles by the VADCP, with the most significant errors in the transverse (cross-flow) velocity direction.

  4. 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)

  5. 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).

  6. 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.

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

    Science.gov (United States)

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

    2017-02-01

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

  8. 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.

  9. 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.

  10. 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

  11. 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.

  12. 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.

  13. 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.

  14. 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)

  15. Axial myopathy

    DEFF Research Database (Denmark)

    Witting, Nanna; Andersen, Linda K; Vissing, John

    2016-01-01

    musculature involvement in the majority of myopathies in which paraspinal musculature was examined. Even in diseases named after a certain pattern of non-axial muscle affection, such as facioscapulohumeral and limb girdle muscular dystrophies, affection of the axial musculature was often severe and early...

  16. Axial-Stereo 3-D Optical Metrology for Inner Profile of Pipes Using a Scanning Laser Endoscope.

    Science.gov (United States)

    Gong, Yuanzheng; Johnston, Richard S; Melville, C David; Seibel, Eric J

    As the rapid progress in the development of optoelectronic components and computational power, 3D optical metrology becomes more and more popular in manufacturing and quality control due to its flexibility and high speed. However, most of the optical metrology methods are limited to external surfaces. This paper proposed a new approach to measure tiny internal 3D surfaces with a scanning fiber endoscope and axial-stereo vision algorithm. A dense, accurate point cloud of internally machined threads was generated to compare with its corresponding X-ray 3D data as ground truth, and the quantification was analyzed by Iterative Closest Points algorithm.

  17. 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.

  18. Influence of impregnation parameters on the axial Mo/. gamma. -alumina profiles studied using a novel simple technique

    Energy Technology Data Exchange (ETDEWEB)

    Goula, M.A.; Kordulis, Ch.; Lycourghiotis, A. (Univ. of Patras (Greece))

    1992-02-01

    The effects of pH, temperature, volume, and concentration of the molybdate solutions, as well as the impregnation time, the rate of drying, the modification of {gamma}-alumina with F{sup {minus}}ions, and the nature and concentration of various competitors, have been systematically studied. It was found that decreases in pH as well as increases in the concentration of the molybdate solutions, of the impregnation temperature, and of the rate of drying cause a progressive transformation of the Mo profile from uniform to eggshell type. Doping of {gamma}-alumina with F{sup {minus}} ions and the use of NH{sub 4}F, H{sub 3}PO{sub 4}, and citric acid as competitors transformed the Mo profiles from eggshell to uniform. The change in volume of the impregnating solution had no effect on the Mo profiles. The effect of the impregnation time was found to be complicated. Increases in the impregnation time, until a critical value sufficient for the complete imbibition of the extrudates, allow for the transfer of great amounts of molybdate to the interior of the extrudates, leading to more uniform profiles. Further increases in the impregnation time caused a considerable increase in the sharpness of the Mo profile. Most of the above observations were explained on the basis of derived equations, adopting a very simple macrodistribution model. Finally, it was demonstrated that small chromatographic columns filled with powder supports may be used to study active ion profiles on catalytic supports.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  20. 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.

  1. 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.

  2. Azimuthal velocity profiles in Rayleigh-stable Taylor–Couette flow and implied axial angular momentum transport

    NARCIS (Netherlands)

    Nordsiek, F.; Huisman, S.G.; Veen, van der R.C.A.; Sun, C.; Lohse, D.; Lathrop, D.P.

    2015-01-01

    We present azimuthal velocity profiles measured in a Taylor–Couette apparatus, which has been used as a model of stellar and planetary accretion disks. The apparatus has a cylinder radius ratio of ${\\it\\eta}=0.716$η=0.716, an aspect ratio of ${\\it\\Gamma}=11.74$Γ=11.74, and the plates closing the cyl

  3. 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)

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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.

  18. 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)

  19. Axially Symmetric, Spatially Homothetic Spacetimes

    CERN Document Server

    Wagh, S M; Wagh, Sanjay M.; Govinder, Keshlan S.

    2002-01-01

    We show that the existence of appropriate spatial homothetic Killing vectors is directly related to the separability of the metric functions for axially symmetric spacetimes. The density profile for such spacetimes is (spatially) arbitrary and admits any equation of state for the matter in the spacetime. When used for studying axisymmetric gravitational collapse, such solutions do not result in a locally naked singularity.

  20. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning; Perfiles axiales de quemado y fraccion de huecos para calculos de criticidad con credito al quemado para combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-07-01

    This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  6. 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)

  7. 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)

  8. 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.

  9. 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.

  10. Light axial vector mesons

    CERN Document Server

    Chen, Kan; Liu, Xiang; Matsuki, Takayuki

    2015-01-01

    Inspired by the abundant experimental observation of axial vector states, we study whether the observed axial vector states can be categorized into the conventional axial vector meson family. In this paper we carry out analysis based on the mass spectra and two-body Okubo-Zweig-Iizuka-allowed decays. Besides testing the possible axial vector meson assignments, we also predict abundant information for their decays and the properties of some missing axial vector mesons, which are valuable to further experimental exploration of the observed and predicted axial vector mesons.

  11. 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.

  12. 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.

  13. 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)

  14. Optimization of axial blowers. Optimierung von Axial-Ventilatoren

    Energy Technology Data Exchange (ETDEWEB)

    Bolte, W.

    1992-08-01

    For the optimum possible design of axial blowers, trials are evaluated in the article, which are based on the grid profile examined by N. Scholz. The computation for the pressure number and the primary degree of efficiency are shown as well as the evaluation of the effect of the Reynolds and mach number on the degree of efficiency and determination of the secondary losses. In a final example, the dimensions of a blower are computed from the data determined during the trials. (orig.).

  15. 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

  16. 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.

  17. 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)

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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).

  5. 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. .

  6. 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.

  7. 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.

  8. 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)

  9. 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)

  10. 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

  11. 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'.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  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. 流化床燃烧超低浓度煤层气的轴向气体分布%Axial Gas Profile During Ultra-Low Concentration Coal Bed Methane Combustion in a Fluidized Bed

    Institute of Scientific and Technical Information of China (English)

    张力; 杨仲卿; 唐强

    2012-01-01

    The ultra-low concentration coal bed methane is difficult to utilize due to its low methane content and fluctuated concentration. The combustion characteristic of ultra-low concentration coal bed methane in a fluidized bed was studied experimentally and numerically. The axial profiles of combustion products were obtained. The effects of inlet methane concentration, bed temperature and fluidized velocity on axial profile were analyzed. The results show that the dimensionless methane concentration decreases with the bed height and reaches a minimum at the bed surface. Then, the value increases abruptly and goes steadily. The CO concentration is less than 20 mL/ma in all experiments. The dimensionless methane concentration at the same bed height decreases with decreasing inlet methane concentration, rising bed temperature and reducing fluidized velocity. Combustion mainly occurred in dense phase, and moved towards lower part of bed with decreasing inlet methane concentration, rising bed temperature and reducing fluidized velocity.%超低浓度煤层气由于甲烷含量低、浓度变化大而较难加以利用。采用实验和数值模拟的方法,研究了超低浓度煤层气在流化床中燃烧特性,得到燃烧产物的轴向分布规律,分析了进气浓度、床层温度、流化风速等因素对甲烷浓度轴向分布的影响。研究结果表明:随着床层高度的增加,无量纲甲烷浓度逐渐减小,在床层表面达到最小值,然后突然增加,随后达到稳定。实验范围内,CO浓度均小于20mL/m3。减小进气浓度、增加床层温度、降低流化风速都会使相同床层高度处的无量纲甲烷浓度减小。燃烧反应主要发生在密相区,随着进气浓度的减小、床层温度的增加、流化风速的降低,反应区域逐渐向床层下部移动。

  18. [Management of axial spondyloarthritis].

    Science.gov (United States)

    Kiltz, U; Baraliakos, X; Braun, J

    2016-11-01

    The term spondyloarthritis (SpA) is now increasingly used to classify and diagnose patients who are characterized by inflammation in the axial skeleton and peripheral manifestations (arthritis and enthesitis). The management of SpA should be tailored according to the current manifestations of the disease, the disease activity and functional impairment. The current article focuses on diagnosis and therapy in patients with axial SpA. Diagnostic procedures are discussed in light of diagnostic utility and feasibility in daily routine care. Cornerstones of treatment in patients with axial SpA are a combination of regular exercise and pharmacological treatment options aiming at anti-inflammatory strategies.

  19. 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.)

  20. 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.

  1. Axial Halbach Magnetic Bearings

    Science.gov (United States)

    Eichenberg, Dennis J.; Gallo, Christopher A.; Thompson, William K.

    2008-01-01

    Axial Halbach magnetic bearings have been investigated as part of an effort to develop increasingly reliable noncontact bearings for future high-speed rotary machines that may be used in such applications as aircraft, industrial, and land-vehicle power systems and in some medical and scientific instrumentation systems. Axial Halbach magnetic bearings are passive in the sense that unlike most other magnetic bearings that have been developed in recent years, they effect stable magnetic levitation without need for complex active control.

  2. 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.

  3. Core burnup calculation and accidents analyses of a pressurized water reactor partially loaded with rock-like oxide fuel

    Science.gov (United States)

    Akie, H.; Sugo, Y.; Okawa, R.

    2003-06-01

    A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.

  4. 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.

  5. 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.

  6. Numerical Simulation on Flow Performance and Blade Profile Optimal Design of Light Load Type Axial Flow Pump%轻载型轴流泵的流场分析和叶型优化

    Institute of Scientific and Technical Information of China (English)

    张宇; 管仁伟; 安伟; 徐峰

    2013-01-01

    在不同工况下,采用商业软件Numeca的Fine/Turbo模块,对包含叶轮、导叶、弯管、喇叭管的轻载型轴流泵进行了全流道的三维湍流数值模拟计算,并在与已有试验数据进行了较好吻合的基础上,对其内部流场进行了详细地数值分析,通过分析轴流泵内部的流动特点,找出了造成原模型流动损失的可能原因,发现由于局部结构设计不合理,流道内产生了漩涡区.针对这一问题提出了改进措施,采用商业软件Numeca的Design3D模块对叶轮叶型压力面和吸力面进行多参数化优化.结果表明,控制压力面、吸力面型线可以有效控制叶片出口处的漩涡,改进后的流道内存在的涡团和流动损失减少,叶轮的水力效和扬程等性能参数相对提高.%Using a commercial software Numeca Fine/Turbo, a numerical algorithm is presented to simulate three-dimensional turbulent flow for a light load type axial flow pump impeller, including the impeller, the guide vane and the bend pipe under different conditions. Through analyzing the axial flow pump internal flow characteristics, we find out the flow loss of the original model, such as vast vortexes areas and wall flow separations. Furthermore, we use the Numeca/Design3D software to optimize the parameters and to redesign the blade profile. The results show that changing the pressure and suctione lines can effectively control the blade of the swirl, resulting in the vortex and flow loss reduction and increasing the efficiency of the impeller hydraulic lift.

  7. 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

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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...

  15. Unsteady Flows in Axial Turbomachines

    Science.gov (United States)

    Marble, F. E.; Rannie, W. D.

    1957-01-01

    Of the various unsteady flows that occur in axial turbomachines certain asymmetric disturbances, of wave length large in comparison with blade spacing, have become understood to a certain extent. These disturbances divide themselves into two categories: self-induced oscillations and force disturbances. A special type of propagating stall appears as a self-induced disturbance; an asymmetric velocity profile introduced at the compressor inlet constitutes a forced disturbance. Both phenomena have been treated from a unified theoretical point of view in which the asymmetric disturbances are linearized and the blade characteristics are assumed quasi-steady. Experimental results are in essential agreement with this theory wherever the limitations of the theory are satisfied. For the self-induced disturbances and the more interesting examples of the forced disturbances, the dominant blade characteristic is the dependence of total pressure loss, rather than the turning angle, upon the local blade inlet angle.

  16. Altered Axial Skeletal Development

    Science.gov (United States)

    The axial skeleton is routinely examined in standard developmental toxicity bioassays and has proven to be sensitive to a wide variety of chemical agents. Dysmorphogenesis in the skull, vertebral column and ribs has been described in both human populations and in laboratory anima...

  17. 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...

  18. 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%.

  19. 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.

  20. 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)

  1. 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.

  2. 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)

  3. 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.

  4. 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)

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

    Energy Technology Data Exchange (ETDEWEB)

    Strain, R V; Johnson, C E

    1978-05-01

    Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

  11. Surface nanoscale axial photonics

    CERN Document Server

    Sumetsky, M

    2011-01-01

    Dense photonic integration promises to revolutionize optical computing and communications. However, efforts towards this goal face unacceptable attenuation of light caused by surface roughness in microscopic devices. Here we address this problem by introducing Surface Nanoscale Axial Photonics (SNAP). The SNAP platform is based on whispering gallery modes circulating around the optical fiber surface and undergoing slow axial propagation readily described by the one-dimensional Schr\\"odinger equation. These modes can be steered with dramatically small nanoscale variation of the fiber radius, which is quite simple to introduce in practice. The extremely low loss of SNAP devices is achieved due to the fantastically low surface roughness inherent in a drawn fiber surface. In excellent agreement with the developed theory, we experimentally demonstrate localization of light in quantum wells, halting light by a point source, tunneling through potential barriers, dark states, etc. This demonstration, prototyping basi...

  12. 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)

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. Critical Axial Load

    Directory of Open Access Journals (Sweden)

    Walt Wells

    2008-01-01

    Full Text Available Our objective in this paper is to solve a second order differential equation for a long, simply supported column member subjected to a lateral axial load using Heun's numerical method. We will use the solution to find the critical load at which the column member will fail due to buckling. We will calculate this load using Euler's derived analytical approach for an exact solution, as well as Euler's Numerical Method. We will then compare the three calculated values to see how much they deviate from one another. During the critical load calculation, it will be necessary to calculate the moment of inertia for the column member.

  2. 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

  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. 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.

  5. 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)

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  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. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. Build Axial Gradient Field by Using Axial Magnetized Permanent Rings

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Axial magnetic field produced by an axial magnetized permanent ring was studied. For two permanent rings, if they are magnetized in the same directions, a nearly uniform axial field can be produced. If they are magnetized in opposite direction,an axial gradient magnetic field can be generated, with the field range changing from -B0 to B0. A permanent magnet with a high axial gradient field was fabricated, the measured results agree with the PANDIRA calculation very well. For wider usage,it is desirable for the field gradient to be changed. Some methods to produce the variable gradient field are presented. These kinds of axial gradient magnetic field can also be used as a beam focusing for linear accelerator if the periodic field can be produced along the beam trajectory. The axial magnetic field is something like a solenoid, large stray field will leak to the outside environment if no method is taken to control them. In this paper, one method is illustrated to shield off the outside leakage field.

  12. Single Band Helical Antenna in Axial Mode

    Directory of Open Access Journals (Sweden)

    Parminder Singh

    2012-11-01

    Full Text Available Helical antennas have been widely used in a various useful applications, due to their low weight and low profile conformability, easy and cheap realization.Radiation properties of this antenna are examined both theoretically and experimentally. In this paper, an attempt has been made to investigate new helical antenna structure for Applications. CST MWS Software is used for the simulation and design calculations of the helical antennas. The axial ratio, return loss, VSWR, Directivity, gain, radiation pattern is evaluated. Using CST MWS simulation software proposed antenna is designed/simulated and optimized. The antenna exhibits a single band from 0 GHz to 3 GHz for GPS and several satellite applications

  13. 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

  14. Bessel beam CARS of axially structured samples

    Science.gov (United States)

    Heuke, Sandro; Zheng, Juanjuan; Akimov, Denis; Heintzmann, Rainer; Schmitt, Michael; Popp, Jürgen

    2015-01-01

    We report about a Bessel beam CARS approach for axial profiling of multi-layer structures. This study presents an experimental implementation for the generation of CARS by Bessel beam excitation using only passive optical elements. Furthermore, an analytical expression is provided describing the generated anti-Stokes field by a homogeneous sample. Based on the concept of coherent transfer functions, the underling resolving power of axially structured geometries is investigated. It is found that through the non-linearity of the CARS process in combination with the folded illumination geometry continuous phase-matching is achieved starting from homogeneous samples up to spatial sample frequencies at twice of the pumping electric field wave. The experimental and analytical findings are modeled by the implementation of the Debye Integral and scalar Green function approach. Finally, the goal of reconstructing an axially layered sample is demonstrated on the basis of the numerically simulated modulus and phase of the anti-Stokes far-field radiation pattern. PMID:26046671

  15. Dissipative Axial Inflation

    CERN Document Server

    Notari, Alessio

    2016-01-01

    We analyze in detail the background cosmological evolution of a scalar field coupled to a massless abelian gauge field through an axial term $\\frac{\\phi}{f_\\gamma} F \\tilde{F}$, such as in the case of an axion. Gauge fields in this case are known to experience tachyonic growth and therefore can backreact on the background as an effective dissipation into radiation energy density $\\rho_R$, which which can lead to inflation without the need of a flat potential. We analyze the system, for momenta $k$ smaller than the cutoff $f_\\gamma$, including numerically the backreaction. We consider the evolution from a given static initial condition and explicitly show that, if $f_\\gamma$ is smaller than the field excursion $\\phi_0$ by about a factor of at least ${\\cal O} (20)$, there is a friction effect which turns on before that the field can fall down and which can then lead to a very long stage of inflation with a generic potential. In addition we find superimposed oscillations, which would get imprinted on any kind of...

  16. Study of axial magnetic effect

    Energy Technology Data Exchange (ETDEWEB)

    Braguta, Victor [IHEP, Protvino, Moscow region, 142284 Russia ITEP, B. Cheremushkinskaya street 25, Moscow, 117218 (Russian Federation); School of Biomedicine, Far Eastern Federal University, Ajax 10 Building 25, Russian island, Vladivostok, 690922 (Russian Federation); Chernodub, M. N. [CNRS, Laboratoire de Mathématiques et Physique Théorique, Université François-Rabelais Tours, Fédération Denis Poisson, Parc de Grandmont, 37200 Tours, France Department of Physics and Astronomy, University of Gent, Krijgslaan 281, S9, B-9000 Gent (Belgium); School of Biomedicine, Far Eastern Federal University, Ajax 10 Building 25, Russian island, Vladivostok, 690922 (Russian Federation); Goy, V. A. [School of Natural Sciences, Far Eastern Federal University, Sukhanova street 8, Vladivostok, 690950 (Russian Federation); Landsteiner, K. [Instituto de Física Teórica UAM/CSIC, C/ Nicolás Cabrera 13-15, Universidad Autónoma de Madrid, Cantoblanco, 28049 Madrid (Spain); Molochkov, A. V. [School of Biomedicine, Far Eastern Federal University, Ajax 10 Building 25, Russian island, Vladivostok, 690922 (Russian Federation); Ulybyshev, M. [ITEP, B. Cheremushkinskaya street 25, Moscow, 117218 Russia Institute for Theoretical Problems of Microphysics, Moscow State University, Moscow, 119899 (Russian Federation)

    2016-01-22

    The Axial Magnetic Effect manifests itself as an equilibrium energy flow of massless fermions induced by the axial (chiral) magnetic field. Here we study the Axial Magnetic Effect in the quenched SU(2) lattice gauge theory with massless overlap fermions at finite temperature. We numerically observe that in the low-temperature hadron phase the effect is absent due to the quark confinement. In the high-temperature deconfinement phase the energy flow is an increasing function of the temperature which reaches the predicted asymptotic T{sup 2} behavior at high temperatures. We find, however, that energy flow is about one order of magnitude lower compared to a theoretical prediction.

  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. 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.

  20. 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).

  1. 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)

  2. Cartilage collagen type II seromarker patterns in axial spondyloarthritis and psoriatic arthritis

    DEFF Research Database (Denmark)

    Munk, Heidi Lausten; Gudmann, Natasja Staehr; Christensen, Anne Friesgaard

    2016-01-01

    The aim of the study was to assess the possible association between type II collagen turnover seromarkers and disease profile in patients with axial spondyloarthritis (SpA) and psoriatic arthritis (PsA). Outpatients with axial SpA (n = 110) or PsA (n = 101) underwent clinical examination including...

  3. Dissipative axial inflation

    Science.gov (United States)

    Notari, Alessio; Tywoniuk, Konrad

    2016-12-01

    We analyze in detail the background cosmological evolution of a scalar field coupled to a massless abelian gauge field through an axial term phi/fγ F ~F, such as in the case of an axion. Gauge fields in this case are known to experience tachyonic growth and therefore can backreact on the background as an effective dissipation into radiation energy density ρR, which can lead to inflation without the need of a flat potential. We analyze the system, for momenta k smaller than the cutoff fγ, including the backreaction numerically. We consider the evolution from a given static initial condition and explicitly show that, if fγ is smaller than the field excursion phi0 by about a factor of at least Script O (20), there is a friction effect which turns on before the field can fall down and which can then lead to a very long stage of inflation with a generic potential. In addition we find superimposed oscillations, which would get imprinted on any kind of perturbations, scalars and tensors. Such oscillations have a period of 4-5 efolds and an amplitude which is typically less than a few percent and decreases linearly with fγ. We also stress that the curvature perturbation on uniform density slices should be sensitive to slow-roll parameters related to ρR rather than dot phi2/2 and we discuss the existence of friction terms acting on the perturbations, although we postpone a calculation of the power spectrum and of non-gaussianity to future work and we simply define and compute suitable slow roll parameters. Finally we stress that this scenario may be realized in the axion case, if the coupling 1/fγ to U(1) (photons) is much larger than the coupling 1/fG to non-abelian gauge fields (gluons), since the latter sets the range of the potential and therefore the maximal allowed phi0~ fG.

  4. 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.

  5. 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.

  6. 基于奇点分布法的轴流泵叶片翼型设计与计算%Design and calculation of airfoil profile of blade in axial flow pump based on singularity approach

    Institute of Scientific and Technical Information of China (English)

    严敬; 刘小兵; 周绪成; 刘小梅; 杨小林

    2016-01-01

    Singularity calculating program is an important approach to design blade airfoils of axial flow machinery. This method is originally used in the runner design of propeller turbines. High efficiency and satisfactory performance of the runners has proved that this program has many advantages compared with other calculating methods for axial flow machines. To improve performance characteristics of axial flow pumps, it is valuable to introduce singularity calculating approach for the design of axial flow pumps. The principles in this program can be described briefly as follows. A vortex sheet is placed along a special curve in the uniform flow field with planar potential flow. If the induced velocity superimposed with the original planar uniform flow can ensure the curve to be a streamline and this streamline can meet all flowing boundary conditions, a solid curved thin plate can be used to replace the vortex sheet, for the flow field formed by the plate and the flow field without the plate are identical. Because velocity distribution of a potential flow is determined by its potential function, which satisfies the Laplacian equation. The solution to any Laplacian equation is solely determined by boundary conditions of the flow. As the induced velocity is developed by vortex sheet, the vortex density distribution along the sheet is very important. In a developed planar flow surface, for the same cascade, energy conversion and relative velocity in the runners and impellers are opposite, and stagnant point and singular point are also located in 2 opposite positions of the same airfoil. As a result, the vortex density distribution along the airfoil mean line can’t be the same for the cascade when used for 2 kinds of hydraulic machines. However, there is only one distribution function presented in traditional approaches reported in all literatures. Further analysis showed that the traditional distribution function was only suitable for boundary conditions of runner

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  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. 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

  15. Instability of Meridional Axial System in f(R) Gravity

    CERN Document Server

    Sharif, M

    2015-01-01

    We analyze dynamical instability of non-static reflection axial stellar structure by taking into account generalized Euler's equation in metric $f(R)$ gravity. Such an equation is obtained by contracting Bianchi identities of usual anisotropic and effective stress-energy tensors, which after using radial perturbation technique gives modified collapse equation. In the realm of $R+\\epsilon R^n$ gravity model, we investigate instability constraints at Newtonian and post-Newtonian approximations. We find that instability of meridional axial self-gravitating system depends upon static profile of structure coefficients while $f(R)$ extra curvature terms induce stability to the evolving celestial body.

  16. Three-dimensional visualization of axial velocity profiles downstream of six different mechanical aortic valve prostheses, measured with a hot-film anemometer in a steady state flow model.

    Science.gov (United States)

    Hasenkam, J M; Westphal, D; Reul, H; Gormsen, J; Giersiepen, M; Stodkilde-Jorgensen, H; Paulsen, P K

    1987-01-01

    Hot-film anemometry was used for in vitro steady-state measurements downstream of six mechanical aortic valve prostheses at flow rates 10, 20 and 30 l.min-1. Three-dimensional visualizations of velocity profiles at two downstream levels were made with the valves rotated 0 and 60 degrees in relation to the sinuses of valsalvae. The velocity fields downstream of the disc valves were generally skew with increasing velocity gradients and laminar shear stresses with increasing flow rates. Furthermore, increased skewness of the velocity profiles was noticed when the major orifices of the disc valves were towards the commissure than when approaching a sinus of valsalvae. The velocity profiles downstream of the ball valve were generally flat but with considerably more disturbed flow, consistent with the findings in turbulent flow.

  17. Relativistic RPA in axial symmetry

    CERN Document Server

    Arteaga, D Pena; 10.1103/PhysRevC.77.034317

    2009-01-01

    Covariant density functional theory, in the framework of self-consistent Relativistic Mean Field (RMF) and Relativistic Random Phase approximation (RPA), is for the first time applied to axially deformed nuclei. The fully self-consistent RMF+RRPA equations are posed for the case of axial symmetry and non-linear energy functionals, and solved with the help of a new parallel code. Formal properties of RPA theory are studied and special care is taken in order to validate the proper decoupling of spurious modes and their influence on the physical response. Sample applications to the magnetic and electric dipole transitions in $^{20}$Ne are presented and analyzed.

  18. Axial Current and Noether Charge

    CERN Document Server

    Mahato, Prasanta

    2012-01-01

    A decade ago, a Lagrangian density has been proposed by the author where only the local symmetries of the Lorentz subgroup of (A)ds group is retained. This formalism has been found to produce some results encompassing that of standard Einstein-Hilbert formalism. In the present article, the conserved axial vector matter currents, constructed in some earlier paper, have been found to be a result of Noether's theorem. PACS: 04.20.Fy, 04.20.Cv, 11.40.-q Keywords: Torsion, Axial Current, Noether's Theorem

  19. 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.

  20. 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...

  1. 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)

  2. Investigation of axial power gradients near a control rod tip

    Energy Technology Data Exchange (ETDEWEB)

    Loberg, John, E-mail: John.Loberg@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Osterlund, Michael, E-mail: Michael.Osterlund@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Bejmer, Klaes-Hakan, E-mail: Klaes-Hakan.Bejmer@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Blomgren, Jan, E-mail: Jan.Blomgren@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Kierkegaard, Jesper, E-mail: Jesper.Kierkegaar@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden)

    2011-07-15

    Highlights: > Pin power gradients near BWR control rod tips have been investigated. > A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. > Small nodes increases pin power gradients; standard nodes underestimates gradients. > The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, {approx}15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

  3. 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%.

  4. 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.

  5. 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 ...

  6. 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.

  7. 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.

  8. 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).

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. Thermophoresis of Axially Symmetric Bodies

    Science.gov (United States)

    2007-11-02

    Sweden Abstract. Thermophoresis of axially symmetric bodies is investigated to first order in the Knudsen-mimber, Kn. The study is made in the limit...derived. Asymptotic solutions are studied. INTRODUCTION Thermophoresis as a phenomenon has been known for a long time, and several authors have approached

  18. Axial structure of the nucleon

    Energy Technology Data Exchange (ETDEWEB)

    Veronique Bernard; Latifa Elouadrhiri; Ulf-G Meissner

    2002-01-01

    We review the current status of experimental and theoretical understanding of the axial nucleon structure at low and moderate energies. Topics considered include (quasi)elastic (anti)neutrino-nucleon scattering, charged pion electroproduction off nucleons and ordinary as well as radiative muon capture on the proton.

  19. 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.

  20. 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.

  1. 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.

  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. 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.

  4. View of the Axial Field Spectrometer

    CERN Multimedia

    1980-01-01

    The Axial Field Spectrometer, with the vertical uranium/scintillator calorimeter and the central drift chamber retracted for service. One coil of the Open Axial Field Magnet is just visible to the right.

  5. Measurement of axial injection displacement with trim coil current unbalance

    Science.gov (United States)

    Covo, Michel Kireeff

    2014-08-01

    The Dee probe used for measuring internal radial beam intensity shows large losses inside the radius of 20 cm of the 88 in. cyclotron. The current of the top and bottom innermost trim coil 1 is unbalanced to study effects of the axial injection displacement. A beam profile monitor images the ion beam bunches, turn by turn. The experimental bunch center of mass position is compared with calculations of the magnetic mirror effect displacement and shows good agreement.

  6. [Axial spondyloarthritis and ankylosing spondylitis].

    Science.gov (United States)

    Nordström, Dan; Kauppi, Markku

    2010-01-01

    Current classification criteria for ankylosing spondylitis do not allow diagnosis before radiographic changes are visible in sacroiliacal joints. The the new axial spondyloarthropathy (SpA) criteria include axial SpA without radiographic changes as well as established ankylosing spondylitis, recognizing them as a continuum of the same disease. This is of major importance as the burden of early SpA is comparable to that of later stage disease. Diagnosis relies on inflammatory MRI findings which is the most significant change compared to earlier criteria. Emerging data on the efficacy of tumor necrosis factor (TNF) alpha blocking therapies already in early but also in established disease have given new promising alternatives for treatment of this often very cumbersome disease, that rarely responds to classic DMARDs.

  7. Axial Spondyloarthritis: An Evolving Concept

    Directory of Open Access Journals (Sweden)

    Nelly Ziadé

    2015-07-01

    Full Text Available Axial spondyloarthritis (AxSpA is the prototype of a family of inter-related yet heterogeneous diseases sharing common clinical and genetic manifestations: the spondyloarthritides (SpAs. The condition mainly affects the sacroiliac joints and axial skeleton, and has a clear classification scheme, wider epidemiological data, and distinct therapeutic guidelines when compared with other SpAs. However, the concept of AxSpA has not been immutable over time and has evolved tremendously on many levels over the past decades. This review identifies the evolution of the AxSpA concept at two levels. First, at the level of classification, the old classifications and rationales leading to the current Assessment of SpondyloArthritis international Society (ASAS classification are reviewed, and the advantages and drawbacks are discussed. Second, at the therapeutic level, current and future treatments are described and treatment strategies are discussed.

  8. 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...

  9. 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)

  10. Golimumab for treatment of axial spondyloarthritis.

    Science.gov (United States)

    Rios Rodriguez, Valeria; Poddubnyy, Denis

    2016-02-01

    Axial spondyloarthritis comprises two forms: nonradiographic (nonradiographic axial spondyloarthritis) and radiographic (better known as ankylosing spondylitis), which are often considered as two stages of one disease. Historically, all currently available TNF-α inhibitors were first investigated in ankylosing spondylitis and later on in nonradiographic axial spondyloarthritis. This year, EMA has granted golimumab approval for the treatment of active nonradiographic axial spondyloarthritis based on the recently published data from the GO-AHEAD study. This article summarizes recent data on efficacy and safety of golimumab in the treatment of ankylosing spondylitis and nonradiographic axial spondyloarthritis.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Etanercept for the treatment of non-radiographic axial spondyloarthritis.

    Science.gov (United States)

    Rios Rodriguez, Valeria; Poddubnyy, Denis

    2016-01-01

    Presently, tumor necrosis factor α antagonist therapy is the only effective alternative treatment to nonsteroidal anti-inflammatory drugs for the entire spectrum of axial spondyloarthritis, including non-radiographic and radiographic (=ankylosing spondylitis) forms. Recently, etanercept has been approved by the European Medicines Agency for the treatment of non-radiographic axial spondyloarthritis, increasing the number of available treatment options for this indication. The latest data on etanercept concerning clinical efficacy and safety in short-term and long-term treatment of patients with non-radiographic axial spondyloarthritis who do not respond to the first-line therapy with non-steroidal anti-inflammatory drugs suggests good efficacy and safety profiles similar to that observed previously in ankylosing spondylitis. This article reviews recent data on the efficacy and safety of etanercept and is focused on the treatment of non-radiographic axial spondyloarthritis. This article will also discuss the role of etanercept in the context of current and developing treatment options.

  2. Axial dispersion in flowing red blood cell suspensions

    Science.gov (United States)

    Podgorski, Thomas; Losserand, Sylvain; Coupier, Gwennou

    2016-11-01

    A key parameter in blood microcirculation is the transit time of red blood cells (RBCs) through an organ, which can influence the efficiency of gas exchange and oxygen availability. A large dispersion of this transit time is observed in vivo and is partly due to the axial dispersion in the flowing suspension. In the classic Taylor-Aris example of a solute flowing in a tube, the combination of molecular diffusion and parabolic velocity profile leads to enhanced axial dispersion. In suspensions of non-Brownian deformable bodies such as RBCs, axial dispersion is governed by a combination of shear induced migration and shear-induced diffusion arising from hydrodynamic interactions. We revisit this problem in the case of RBC pulses flowing in a microchannel and show that the axial dispersion of the pulse eventually saturates with a final extension that depends directly on RBC mechanical properties. The result is especially interesting in the dilute limit since the final pulse length depends only on the channel width, exponent of the migration law and dimensionless migration velocity. In continuous flow, the dispersion of transit times is the result of complex cell-cell and cell-wall interactions and is strongy influenced by the polydispersity of the blood sample. The authors acknowledge support from LabEx TEC21 and CNES.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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...

  8. 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.)

  9. 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.

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-01

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

  12. 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.

  13. 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 ...

  14. 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.

  15. 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

  16. Axial Globe Length in Congenital Ptosis.

    Science.gov (United States)

    Takahashi, Yasuhiro; Kang, Hyera; Kakizaki, Hirohiko

    2015-01-01

    To compare axial globe length between affected and unaffected sides in patients with unilateral congenital ptosis. This prospective observational study included 37 patients (age range: 7 months to 58 years). The axial globe length, margin reflex distance-1 (MRD-1), and refractive power were measured. The axial globe length difference was calculated by subtracting the axial globe length on the unaffected side from that of the affected side. The relationships among axial globe length differences, MRD-1 on the affected sides, and patient ages were analyzed using multiple regression analysis. No significant differences were found in the axial globe length between sides (P = .677). The axial globe length difference was 0.17 ± 0.30 mm (mean ± standard deviation), and two patients (5.4%), aged 32 to 57 years, showed axial globe length more than 0.67 mm longer (corresponding to a refractive power of 2 diopters) on the affected side compared to the unaffected side. The multiple regression model between axial globe length difference, patient age, and MRD-1 on the affected sides was less appropriate (YAGL = 0.003XAGE-0.048XMRD-1 +0.112; r = 0.338; adjusted r2 = 0.062; P = .127). The cylindrical power was greater on the affected side (P = .046), although the spherical power was not different between sides (P = .657). No significant difference was identified in the axial globe length between sides, and only 5% of non-pediatric patients showed an axial globe length more than 0.67 mm longer on the affected side. Congenital ptosis may have little effect on axial globe length elongation, and the risk of axial myopia-induced anisometropic amblyopia may be low in patients with unilateral congenital ptosis. Copyright 2015, SLACK Incorporated.

  17. 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

  18. Axial Vector $Z'$ and Anomaly Cancellation

    CERN Document Server

    Ismail, Ahmed; Tsao, Kuo-Hsing; Unwin, James

    2016-01-01

    Whilst the prospect of new $Z'$ gauge bosons with only axial couplings to the Standard Model (SM) fermions is widely discussed, examples of anomaly-free renormalisable models are lacking in the literature. We look to remedy this by constructing several motivated examples. Specifically, we consider axial vectors which couple universally to all SM fermions, as well as those which are generation-specific, leptophilic, and leptophobic. Anomaly cancellation typically requires the presence of new coloured and charged chiral fermions, and we argue that the masses of these new states must generally be comparable to that of the axial vector. Finally, an axial vector mediator could provide a portal between SM and hidden sector states, and we also consider the possibility that the axial vector couples to dark matter. If the dark matter relic density is set due to freeze-out via the axial vector, this strongly constrains the parameter space.

  19. Axial vector Z‧ and anomaly cancellation

    Science.gov (United States)

    Ismail, Ahmed; Keung, Wai-Yee; Tsao, Kuo-Hsing; Unwin, James

    2017-05-01

    Whilst the prospect of new Z‧ gauge bosons with only axial couplings to the Standard Model (SM) fermions is widely discussed, examples of anomaly-free renormalisable models are lacking in the literature. We look to remedy this by constructing several motivated examples. Specifically, we consider axial vectors which couple universally to all SM fermions, as well as those which are generation-specific, leptophilic, and leptophobic. Anomaly cancellation typically requires the presence of new coloured and charged chiral fermions, and we argue that in a large class of models masses of these new states are expected to be comparable to that of the axial vector. Finally, an axial vector mediator could provide a portal between SM and hidden sector states, and we also consider the possibility that the axial vector couples to dark matter. If the dark matter relic density is set due to freeze-out via the axial vector, this strongly constrains the parameter space.

  20. Instability of meridional axial system in f(R) gravity

    Energy Technology Data Exchange (ETDEWEB)

    Sharif, M.; Yousaf, Z. [University of the Punjab, Department of Mathematics, Lahore (Pakistan)

    2015-05-15

    We analyze the dynamical instability of a non-static reflection axial stellar structure by taking into account the generalized Euler equation in metric f(R) gravity. Such an equation is obtained by contracting the Bianchi identities of the usual anisotropic and effective stress-energy tensors, which after using a radial perturbation technique gives a modified collapse equation. In the realm of the R + εR{sup n} gravity model, we investigate instability constraints at Newtonian and post-Newtonian approximations. We find that the instability of a meridional axial self-gravitating system depends upon the static profile of the structure coefficients, while f(R) extra curvature terms induce the stability of the evolving celestial body. (orig.)

  1. Diabetes mellitus and the eye: axial length

    OpenAIRE

    Huntjens, B.; O’Donnell, C.

    2006-01-01

    Background and aims: The refractive error of the eye is dependent on its axial length. Refractive error is known to fluctuate significantly in poorly controlled diabetic patients. Recently it has been reported that human eyes fluctuate in axial length during the day. However, this change is not detectable in all subjects, suggesting physiological influences such as diet. The purpose of this study was to investigate fluctuations in axial length and blood glucose levels (BGLs) in diabetic patie...

  2. System Study for Axial Vane Engine Technology

    Science.gov (United States)

    Badley, Patrick R.; Smith, Michael R.; Gould, Cedric O.

    2008-01-01

    The purpose of this engine feasibility study was to determine the benefits that can be achieved by incorporating positive displacement axial vane compression and expansion stages into high bypass turbofan engines. These positive-displacement stages would replace some or all of the conventional compressor and turbine stages in the turbine engine, but not the fan. The study considered combustion occurring internal to an axial vane component (i.e., Diesel engine replacing the standard turbine engine combustor, burner, and turbine); and external continuous flow combustion with an axial vane compressor and an axial vane turbine replacing conventional compressor and turbine systems.

  3. The Emergence of Axial Parts

    Directory of Open Access Journals (Sweden)

    Peter Svenonius

    2007-01-01

    Full Text Available Many languages have specialized locative words or morphemes translating roughly into words like ‘front,’ ‘back,’ ‘top,’ ‘bottom,’ ‘side,’ and so on. Often, these words are used instead of more specialized adpositions to express spatial meanings corresponding to ‘behind,’ ‘above,’ and so on. I argue, on the basis of a cross-linguistic survey of such expressions, that in many cases they motivate a syntactic category which is distinct from both N and P, which I call AxPart for ‘Axial Part’; I show how the category relates to the words which instantiate it, and how the meaning of the construction is derived from the combination of P[lace] elements, AxParts, and the lexical material which expresses them.

  4. 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.

  5. 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.

  6. 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

  7. 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.

  8. Health and imaging outcomes in axial spondyloarthritis

    NARCIS (Netherlands)

    Machado, P.M.

    2016-01-01

    This thesis focuses on the assessment and monitoring of health and imaging outcomes in axial spondyloarthritis (SpA) and the relationship between these outcomes. Four major contributions to the understanding and management of axial SpA were made: 1) the improvement and facilitation of the assessment

  9. 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.

  10. 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

  11. Universal Axial Fluctuations in Optical Tweezers

    CERN Document Server

    Ribezzi-Crivellari, Marco; Ritort, Felix

    2015-01-01

    Optical tweezers allow the measurement of fluctuations at the nano-scale, in particular fluctuations in the end-to-end distance in single molecules. Fluctuation spectra can yield valuable information, but they can easily be contaminated by instrumental effects. We identify axial fluctuations, i.e. fluctuations of the trapped beads in the direction of light propagation, as one of these instrumental effects. Remarkably, axial fluctuations occur on a characteristic timescale similar to that of conformational (folding) transitions, which may lead to misinterpretation of the experimental results. We show that a precise measurement of the effect of force on both axial and conformational fluctuations is crucial to disentangle them. Our results on axial fluctuations are captured by a simple and general formula valid for all optical tweezers setups and provide experimentalists with a general strategy to distinguish axial fluctuations from conformational transitions.

  12. Novel Integration Radial and Axial Magnetic Bearing

    Science.gov (United States)

    Blumenstock, Kenneth; Brown, Gary

    2000-01-01

    Typically, fully active magnetically suspended systems require one axial and two radial magnetic bearings. Combining radial and axial functions into a single device allows for more compact and elegant packaging. Furthermore, in the case of high-speed devices such as energy storage flywheels, it is beneficial to minimize shaft length to keep rotor mode frequencies as high as possible. Attempts have been made to combine radial and axial functionality, but with certain drawbacks. One approach requires magnetic control flux to flow through a bias magnet reducing control effectiveness, thus resulting in increased resistive losses. This approach also requires axial force producing magnetic flux to flow in a direction into the rotor laminate that is undesirable for minimizing eddy-current losses resulting in rotational losses. Another approach applies a conical rotor shape to what otherwise would be a radial heteropolar magnetic bearing configuration. However, positional non-linear effects are introduced with this scheme and the same windings are used for bias, radial, and axial control adding complexity to the controller and electronics. For this approach, the amount of axial capability must be limited. It would be desirable for an integrated radial and axial magnetic bearing to have the following characteristics; separate inputs for radial and axial control for electronics and control simplicity, all magnetic control fluxes should only flow through their respective air gaps and should not flow through any bias magnets for minimal resistive losses, be of a homopolar design to minimize rotational losses, position related non-linear effects should be minimized, and dependent upon the design parameters, be able to achieve any radial/axial force or power ratio as desired. The integrated radial and axial magnetic bearing described in this paper exhibits all these characteristics. Magnetic circuit design, design equations, and magnetic field modeling results will be presented.

  13. Novel Integrated Radial and Axial Magnetic Bearing

    Science.gov (United States)

    Blumenstock, Kenneth A.; Brown, Gary L.; Powers, Edward I. (Technical Monitor)

    2000-01-01

    Typically, fully active magnetically suspended systems require one axial and two radial magnetic bearings. Combining radial and axial functions into a single device allows for more compact and elegant packaging. Furthermore, in the case of high-speed devices such as energy storage flywheels, it is beneficial to minimize shaft length to keep rotor mode frequencies as high as possible. Attempts have been made to combine radial and axial functionality, but with certain drawbacks. One approach requires magnetic control flux to flow through a bias magnet reducing control effectiveness, thus resulting in increased resistive losses. This approach also requires axial force producing magnetic flux to flow in a direction into the rotor laminate that is undesirable for minimizing eddy-current losses resulting in rotational losses. Another approach applies a conical rotor shape to what otherwise would be a radial heteropolar magnetic bearing configuration. However, positional non-linear effects are introduced with this scheme and the same windings are used for bias, radial, and axial control adding complexity to the controller and electronics. For this approach, the amount of axial capability must be limited. It would be desirable for an integrated radial and axial magnetic bearing to have the following characteristics, separate inputs for radial and axial control for electronics and control simplicity, all magnetic control fluxes should only flow through their respective air gaps and should not flow through any bias magnets for minimal resistive losses, be of a homopolar design to minimize rotational losses, position related non-linear effects should be minimized, and dependent upon the design parameters, be able to achieve any radial/axial force or power ratio as desired. The integrated radial and axial magnetic bearing described in this paper exhibits all these characteristics. Magnetic circuit design, design equations, and analysis results will be presented.

  14. Axial design of nuclear fuel using path relinking; Diseno axial de combustible nuclear utilizando path relinking

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, A.; Torres, M.; Ortiz, J. J.; Perusquia, R.; Hernandez, J. L.; Montes, J. L. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jacm@nuclear.inin.mx

    2008-07-01

    In the present work the preliminary results were obtained with the zoctli system whose purpose is the axial design of assembly of nuclear fuel under certain considerations. For the mentioned design well-know cells were already used and that they have been proven in diverse cycles of operation in the nuclear power plant of Laguna Verde. The design contemplates fuels assemblies of 10x10 and with 2 water channels. The assembly was distributed in 6 axial zones according to its structure. In order to take to end the optimization is was used the well-known technique like Path relinking and to find the group of previous solutions required by this technique uses the technical Taboo search. In order to work with Path relinking, 5 trajectories was taken in to account from a set of 5 previous solutions generated with theTaboo search, the update of the group of solutions is carried out in dynamic form. In the case of the Taboo search it was used a list of variable size, it was implement an aspiration approach, it was used the vector of frequencies and due to the cost of the evaluation of the objective function, only it was review 5% of the vicinity. For the objective function was considered the limit thermal, the axial profile of power, the effective multiplication factor and the margin of having turned off in cold. In order to prove the design system, it was used a balance cycle with a value of reference of 0.9928 for the effective multiplication factor that is equivalent to a produced energy of 10896 MWd/TU at the end of operation to full power. The designed assemblies were placed both in one of lots different from fresh assemblies on which it counts the referred cycle. At the end one a comparison with the results obtained with other techniques and under similar conditions is made. The results obtained until the moment show an appropriate performance of the system. It is possible to indicate that a small inconvenient is the amount of consumed resources of calculation during

  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. 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

  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. Axial stress localization facilitates pressure propagation in gelled pipes

    Science.gov (United States)

    Norrman, Jens; Skjæraasen, Olaf; Oschmann, Hans-Jörg; Paso, Kristofer; Sjöblom, Johan

    2016-03-01

    Paraffin wax-oil gels are unique rheological fluids which undergo shear degradation starting at a deformation (shear strain level) of approximately 1%. Flow commencement in pipelines filled with wax-oil gels is a complex hydrodynamic process involving propagation of acoustic, diffusive, and rheological degradation pressure wave fronts. Dynamic simulation informed by qualified rheological relations provides useful insight into the physical nature of these flow processes. Eulerian simulations are presented which emulate known physical phenomena and essential characteristics of wax-gel flow dynamics. A constitutive rheological equation set accounts for deformation-driven reduction in yield stress and viscosity terms. No explicit time-dependent rheological parameters are utilized in the equations. Rheological yielding alters the nature of the dominant pressure wave from inherently diffusive towards self-sharpening. Axial stress localization effectively sequentializes the gel breakage process, quantified by reduced length of the pressure wave-front zone. Ultimately, axial stress localization allows flow in longer pipe segments, albeit with a concomitant time delay. Viscous behavior and yielding degradation behavior are shown to account for upward and downward concavity in transient axial pressure profiles, respectively. Overall, a unique synergy between gel compressibility and gel degradation is revealed. Deformation-coupled interaction between compressibility and degradation allows pressure propagation and subsequent sustained flow through a gel material which is otherwise immobile in the incompressible case.

  19. Axial force measurement for esophageal function testing

    Institute of Scientific and Technical Information of China (English)

    Flemming H Gravesen; Peter Funch-Jensen; Hans Gregersen; Asbjφrn Mohr Drewes

    2009-01-01

    The esophagus serves to transport food and fluid from the pharynx to the stomach. Manometry has been the "golden standard" for the diagnosis of esophageal motility diseases for many decades. Hence, esophageal function is normally evaluated by means of manometry even though it reflects the squeeze force (force in radial direction) whereas the bolus moves along the length of esophagus in a distal direction. Force measurements in the longitudinal (axial) direction provide a more direct measure of esophageal transport function. The technique used to record axial force has developed from external force transducers over in-vivo strain gauges of various sizes to electrical impedance based measurements. The amplitude and duration of the axial force has been shown to be as reliable as manometry. Normal, as well as abnormal, manometric recordings occur with normal bolus transit, which have been documented using imaging modalities such as radiography and scintigraphy. This inconsistency using manometry has also been documented by axial force recordings. This underlines the lack of information when diagnostics are based on manometry alone. Increasing the volume of a bag mounted on a probe with combined axial force and manometry recordings showed that axial force amplitude increased by 130% in contrast to an increase of 30% using manometry. Using axial force in combination with manometry provides a more complete picture of esophageal motility, and the current paper outlines the advantages of using this method.

  20. Axial force measurement for esophageal function testing.

    Science.gov (United States)

    Gravesen, Flemming H; Funch-Jensen, Peter; Gregersen, Hans; Drewes, Asbjørn Mohr

    2009-01-14

    The esophagus serves to transport food and fluid from the pharynx to the stomach. Manometry has been the "golden standard" for the diagnosis of esophageal motility diseases for many decades. Hence, esophageal function is normally evaluated by means of manometry even though it reflects the squeeze force (force in radial direction) whereas the bolus moves along the length of esophagus in a distal direction. Force measurements in the longitudinal (axial) direction provide a more direct measure of esophageal transport function. The technique used to record axial force has developed from external force transducers over in-vivo strain gauges of various sizes to electrical impedance based measurements. The amplitude and duration of the axial force has been shown to be as reliable as manometry. Normal, as well as abnormal, manometric recordings occur with normal bolus transit, which have been documented using imaging modalities such as radiography and scintigraphy. This inconsistency using manometry has also been documented by axial force recordings. This underlines the lack of information when diagnostics are based on manometry alone. Increasing the volume of a bag mounted on a probe with combined axial force and manometry recordings showed that axial force amplitude increased by 130% in contrast to an increase of 30% using manometry. Using axial force in combination with manometry provides a more complete picture of esophageal motility, and the current paper outlines the advantages of using this method.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. New Anomaly of the Axial-Vector Current

    Institute of Scientific and Technical Information of China (English)

    HE Han-Xin

    2001-01-01

    By computing the axial-vector current operator equation, we find the anomalous axial-vector curl equation besides the well-known anomalous axial-vector divergence equation (the Adler-Bell-Jackiw anomaly) and discuss its implication.``

  7. Comparison of Design Methods for Axially Loaded Driven Piles in Cohesionless Soil

    DEFF Research Database (Denmark)

    Thomassen, Kristina; Andersen, Lars Vabbersgaard; Ibsen, Lars Bo

    2012-01-01

    For offshore wind turbines on deeper waters, a jacket sub-structure supported by axially loaded piles is thought to be the most suitable solution. The design method recommended by API and two CPT-based design methods are compared for two uniform sand profiles. The analysis show great difference...

  8. Axial force measurement for esophageal function testing

    DEFF Research Database (Denmark)

    Gravesen, Flemming Holbæk; Funch-Jensen, Peter; Gregersen, Hans

    2009-01-01

    force transducers over in-vivo strain gauges of various sizes to electrical impedance based measurements. The amplitude and duration of the axial force has been shown to be as reliable as manometry. Normal, as well as abnormal, manometric recordings occur with normal bolus transit, which have been...... force (force in radial direction) whereas the bolus moves along the length of esophagus in a distal direction. Force measurements in the longitudinal (axial) direction provide a more direct measure of esophageal transport function. The technique used to record axial force has developed from external...... documented using imaging modalities such as radiography and scintigraphy. This inconsistency using manometry has also been documented by axial force recordings. This underlines the lack of information when diagnostics are based on manometry alone. Increasing the volume of a bag mounted on a probe...

  9. Axial Super-resolution Evanescent Wave Tomography

    CERN Document Server

    Pendharker, Sarang; Newman, Ward; Ogg, Stephen; Nazemifard, Neda; Jacob, Zubin

    2016-01-01

    Optical tomographic reconstruction of a 3D nanoscale specimen is hindered by the axial diffraction limit, which is 2-3 times worse than the focal plane resolution. We propose and experimentally demonstrate an axial super-resolution evanescent wave tomography (AxSET) method that enables the use of regular evanescent wave microscopes like Total Internal Reflection Fluorescence Microscope (TIRF) beyond surface imaging, and achieve tomographic reconstruction with axial super-resolution. Our proposed method based on Fourier reconstruction achieves axial super-resolution by extracting information from multiple sets of three-dimensional fluorescence images when the sample is illuminated by an evanescent wave. We propose a procedure to extract super-resolution features from the incremental penetration of an evanescent wave and support our theory by 1D (along the optical axis) and 3D simulations. We validate our claims by experimentally demonstrating tomographic reconstruction of microtubules in HeLa cells with an axi...

  10. How to diagnose axial spondyloarthritis early

    OpenAIRE

    Rudwaleit, M.; van der Heijde, D.; Khan, M.; Braun, J.; Sieper, J.

    2004-01-01

    Background: Chronic low back pain (LBP), the leading symptom of ankylosing spondylitis (AS) and undifferentiated axial spondyloarthritis (SpA), precedes the development of radiographic sacroiliitis, sometimes by many years.

  11. Axial thermal rotation of slender rods.

    Science.gov (United States)

    Li, Dichuan; Fakhri, Nikta; Pasquali, Matteo; Biswal, Sibani Lisa

    2011-05-06

    Axial rotational diffusion of rodlike polymers is important in processes such as microtubule filament sliding and flagella beating. By imaging the motion of small kinks along the backbone of chains of DNA-linked colloids, we produce a direct and systematic measurement of axial rotational diffusivity of rods both in bulk solution and near a wall. The measured diffusivities decrease linearly with the chain length, irrespective of the distance from a wall, in agreement with slender-body hydrodynamics theory. Moreover, the presence of small kinks does not affect the chain's axial diffusivity. Our system and measurements provide insights into fundamental axial diffusion processes of slender objects, which encompass a wide range of entities including biological filaments and linear polymer chains.

  12. Axial light emission and Ar metastable densities in a parallel plate dc micro discharge in steady state and transient regimes

    CERN Document Server

    Kuschel, T; Stefanović, I; Böke, M; Skoro, N; Marić, D; Petrović, Z Lj; Winter, J

    2011-01-01

    Axial emission profiles in a parallel plate dc micro discharge (feedgas: argon; discharge gap d=1mm; pressure p=10Torr) were studied by means of time resolved imaging with a fast ICCD camera. Additionally, volt-ampere (V-A) characteristics were recorded and Ar* metastable densities were measured by tunable diode laser absorption spectroscopy (TDLAS). Axial emission profiles in the steady state regime are similar to corresponding profiles in standard size discharges (d=1cm, p=1Torr). For some discharge conditions relaxation oscillations are present when the micro discharge switches periodically between low current Townsend-like mode and normal glow. At the same time the axial emission profile shows transient behavior, starting with peak distribution at the anode, which gradually moves towards the cathode during the normal glow. The development of argon metastable densities highly correlates with the oscillating discharge current. Gas temperatures in the low current Townsend-like mode (T= 320-400K) and the high...

  13. Nonperturbative Aspects of Axial Vector Vertex

    Institute of Scientific and Technical Information of China (English)

    ZONG Hong-Shi; CHEN Xiang-Song; WANG Fan; CHANG Chao-Hsi; ZHAO En-Guang

    2002-01-01

    It is shown how the axial vector current of current quarks is related to that of constituent quarks within the framework of the global color symmetry model.Gluon dressing of the axial vector vertex and the quark self-energy functions are described by the inhomogeneous Bethe-Salpeter equation in the ladder approximation and the Schwinger Dyson equation in the rainbow approximation,respectively.

  14. Numerical simulation of axial flow compressors.

    OpenAIRE

    Jesuino Takachi Tomita

    2002-01-01

    This work deals with the numerical simulation of axial flow compressors, from design to performance prediction. The stage performance prediction uses the meanline flow properties. Stage-stacking is used to analyse a multi-stage compressor. A computer program, written in FORTRAN, was developed and is able to design an axial flow compressor given air mass flow, total pressure ratio, overall efficiency and design speed. All geometrical data relevant to the compressor performance prediction is ca...

  15. Wave propagation in axially moving periodic strings

    Science.gov (United States)

    Sorokin, Vladislav S.; Thomsen, Jon Juel

    2017-04-01

    The paper deals with analytically studying transverse waves propagation in an axially moving string with periodically modulated cross section. The structure effectively models various relevant technological systems, e.g. belts, thread lines, band saws, etc., and, in particular, roller chain drives for diesel engines by capturing both their spatial periodicity and axial motion. The Method of Varying Amplitudes is employed in the analysis. It is shown that the compound wave traveling in the axially moving periodic string comprises many components with different frequencies and wavenumbers. This is in contrast to non-moving periodic structures, for which all components of the corresponding compound wave feature the same frequency. Due to this "multi-frequency" character of the wave motion, the conventional notion of frequency band-gaps appears to be not applicable for the moving periodic strings. Thus, for such structures, by frequency band-gaps it is proposed to understand frequency ranges in which the primary component of the compound wave attenuates. Such frequency band-gaps can be present for a moving periodic string, but only if its axial velocity is lower than the transverse wave speed, and, the higher the axial velocity, the narrower the frequency band-gaps. The revealed effects could be of potential importance for applications, e.g. they indicate that due to spatial inhomogeneity, oscillations of axially moving periodic chains always involve a multitude of frequencies.

  16. An Unbroken Axial Vector Current Conservation Law

    CERN Document Server

    Sharafiddinov, Rasulkhozha S

    2015-01-01

    The mass, energy and momentum of the neutrino of a true flavor have an axial-vector nature. As a consequence, the left-handed truly neutral neutrino in an axial-vector field of emission can be converted into a right-handed one and vice versa. This predicts the unidenticality of masses, energies and momenta of neutrinos of the different components. Recognizing such a difference in masses, energies, momenta and accepting that the left-handed axial-vector neutrino and the right-handed antineutrino of true neutrality refer to long-lived C-odd leptons, and the right-handed truly neutral neutrino and the left-handed axial-vector antineutrino are of short-lived fermions of C-oddity, we would write a new CP-even Dirac equation taking into account the flavor symmetrical axial-vector mass, energy and momentum matrices. Their presence explains the spontaneous mirror symmetry violation, confirming that an axial-vector current conservation law has never violated. They reflect the availability of a mirror Minkowski space i...

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

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

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

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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)

  3. Analytical and numerical investigation of the optimum pressure distribution along a low-pressure axial fan blade

    OpenAIRE

    2013-01-01

    In the field of axial flow turbomachines, the two-dimensional cascade model is often used experimentally or numerically to investigate fundamental flow characteristics and overall performance of the impeller. The core of the present work is a design method for axial fan cascades aiming to derive inversely the optimum blade shape based on the requirements of the impeller and not using any predefined aerofoil profiles. While most design strategies based on the aerofoil theory assume constant to...

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. Comparison of Design Methods for Axially Loaded Driven Piles in Cohesionless Soil

    DEFF Research Database (Denmark)

    Thomassen, Kristina; Andersen, Lars Vabbersgaard; Ibsen, Lars Bo

    2012-01-01

    For offshore wind turbines on deeper waters, a jacket sub-structure supported by axially loaded piles is thought to be the most suitable solution. The design method recommended by API and two CPT-based design methods are compared for two uniform sand profiles. The analysis show great difference...... in the predictions of bearing capacities calculated by means of the three methods for piles loaded in both tension and compression. This implies that further analysis of the bearing capacity of axially loaded piles in sand should be conducted....

  10. Aerodynamics and combustion of axial swirlers

    Science.gov (United States)

    Fu, Yongqiang

    A multipoint lean direct injection (LDI) concept was introduced recently in non-premixed combustion to obtain both low NOx emissions and good combustion stability. In this concept, a key feature is the injection of finely atomized fuel into the high-swirling airflow at the combustor dome that provides a homogenous, lean fuel-air mixture. In order to achieve the fine atomization and mixing of the fuel and air quickly and uniformly, a good swirler design should be studied. The focus of this dissertation is to investigate the aerodynamics and combustion of the swirling flow field in a multipoint Lean Direct Injector combustor. A helical axial-vaned swirler with a short internal convergent-divergent venturi was used. Swirlers with various vane angles and fuel nozzle insertion lengths have been designed. Three non-dimensional parameter effects on non-reacting, swirling flow field were studied: swirler number, confinement ratio and Reynolds number. Spray and combustion characteristics on the single swirler were studied to understand the mechanism of fuel-air mixing in this special configuration. Multi-swirler interactions were studied by measuring the confined flow field of a multipoint swirler array with different configurations. Two different swirler arrangements were investigated experimentally, which include a co-swirling array and a counter-swirling array. In order to increase the range of stability of multipoint LDI combustors, an improved design were also conducted. The results show that the degree of swirl and the level of confinement have a clear impact on the mean and turbulent flow fields. The swirling flow fields may also change significantly with the addition of a variety of simulated fuel nozzle insertion lengths. The swirler with short insertion has the stronger swirling flow as compared with the long insertion swirler. Reynolds numbers, with range of current study, will not alter mean and turbulent properties of generated flows. The reaction of the spray

  11. Optimization of residual heat removal pump axial thrust and axial bearing

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, F.

    1996-12-01

    The residual heat removal (RHR) pumps of German 1300 megawatt pressurized-water reactor (PWR) power plants are of the single stage end suction type with volute casing or with diffuser and forged circular casing. Due to the service conditions the pumps have to cover the full capacity range as well as a big variation in suction static pressure. This results in a big difference in the axial thrust that has to be borne by the axial bearing. Because these pumps are designed to operate without auxiliary systems (things that do not exist can not fail), they are equipped with antifriction bearings and sump oil lubrication. To minimize the heat production within the bearing casing, a number of PWR plants have pumps with combined axial/radial bearings of the ball type. Due to the fact that the maximum axial thrust caused by static pressure and hydrodynamic forces on the impeller is too big to be borne by that type of axial bearing, the impellers were designed to produce a hydrodynamic axial force that counteracts the static axial force. Thus, the resulting axial thrust may change direction when the static pressure varies.

  12. Reducing axial mixing in flotation columns

    Energy Technology Data Exchange (ETDEWEB)

    Al Taweel, A.M.; Ramadan, A.M. [Technical Univ. of Nova Scotia, Halifax (Canada). Chemical Engineering Dept.; Moharam, M.R.; Hassan, T.A. [Al Azhar Univ., Cairo (Egypt); El Mofty, S.M. [Cairo Univ., Giza (Egypt)

    1995-10-01

    The axial mixing characteristics of a pilot-scale flotation column were investigated with the objective of identifying means to mitigate the extent of axial mixing that adversely affects its grade/recovery performance. A wide range of design and operating conditions wa investigated and the experimental results, obtained using the dynamic response method, were analyzed using three axial mixing models. The dynamic response of the column can best be described using the axial dispersion model. The results obtained suggest that the value of the axial dispersion coefficient, E{sub L}, can be significantly reduced by judicial selection of hydrodynamic conditions and/or the use of column inserts that suppress the onset of hydrodynamic instabilities inherent to the operation of conventional flotation columns. Up to 40% reduction in the value of E{sub L} was thus obtained by using spargers that produce more uniform bubble sizes, while up to 30% reductions were obtained by controlling the residual frother concentration. 33 refs., 7 figs.

  13. Axial super-resolution evanescent wave tomography.

    Science.gov (United States)

    Pendharker, Sarang; Shende, Swapnali; Newman, Ward; Ogg, Stephen; Nazemifard, Neda; Jacob, Zubin

    2016-12-01

    Optical tomographic reconstruction of a three-dimensional (3D) nanoscale specimen is hindered by the axial diffraction limit, which is 2-3 times worse than the focal plane resolution. We propose and experimentally demonstrate an axial super-resolution evanescent wave tomography method that enables the use of regular evanescent wave microscopes like the total internal reflection fluorescence microscope beyond surface imaging and achieve a tomographic reconstruction with axial super-resolution. Our proposed method based on Fourier reconstruction achieves axial super-resolution by extracting information from multiple sets of 3D fluorescence images when the sample is illuminated by an evanescent wave. We propose a procedure to extract super-resolution features from the incremental penetration of an evanescent wave and support our theory by one-dimensional (along the optical axis) and 3D simulations. We validate our claims by experimentally demonstrating tomographic reconstruction of microtubules in HeLa cells with an axial resolution of ∼130  nm. Our method does not require any additional optical components or sample preparation. The proposed method can be combined with focal plane super-resolution techniques like stochastic optical reconstruction microscopy and can also be adapted for THz and microwave near-field tomography.

  14. Axial super-resolution evanescent wave tomography

    Science.gov (United States)

    Pendharker, Sarang; Shende, Swapnali; Newman, Ward; Ogg, Stephen; Nazemifard, Neda; Jacob, Zubin

    2016-12-01

    Optical tomographic reconstruction of a 3D nanoscale specimen is hindered by the axial diffraction limit, which is 2-3 times worse than the focal plane resolution. We propose and experimentally demonstrate an axial super-resolution evanescent wave tomography (AxSET) method that enables the use of regular evanescent wave microscopes like Total Internal Reflection Fluorescence Microscope (TIRF) beyond surface imaging, and achieve tomographic reconstruction with axial super-resolution. Our proposed method based on Fourier reconstruction achieves axial super-resolution by extracting information from multiple sets of three-dimensional fluorescence images when the sample is illuminated by an evanescent wave. We propose a procedure to extract super-resolution features from the incremental penetration of an evanescent wave and support our theory by 1D (along the optical axis) and 3D simulations. We validate our claims by experimentally demonstrating tomographic reconstruction of microtubules in HeLa cells with an axial resolution of $\\sim$130 nm. Our method does not require any additional optical components or sample preparation. The proposed method can be combined with focal plane super-resolution techniques like STORM and can also be adapted for THz and microwave near-field tomography.

  15. Atlanto-axial infection after acupuncture.

    Science.gov (United States)

    Robinson, A; Lind, C R P; Smith, R J; Kodali, V

    2015-12-11

    A 67-year-old man presented with neck cellulitis following acupuncture for cervical spondylosis. Blood cultures were positive for methicillin-sensitive Staphylococcus aureus. Increased neck pain and bacteraemia prompted MRI, which showed atlanto-axial septic arthritis without signs of infection of the tissues between the superficial cellulitic area and the atlanto-axial joint, thus making direct extension of infection unlikely. It is more likely that haematogenous spread of infection resulted in seeding in the atlanto-axial joint, with the proximity of the arthritis and acupuncture site being coincidental. Acupuncture is a treatment option for some indolent pain conditions. As such, acupuncture services are likely to be more frequently utilised. A history of acupuncture is rarely requested by the admitting doctor and seldom offered voluntarily by the patient, especially where the site of infection due to haematogenous spread is distant from the needling location. Awareness of infectious complications following acupuncture can reduce morbidity through early intervention.

  16. Axial symmetry and conformal Killing vectors

    CERN Document Server

    Mars, M; Mars, Marc; Senovilla, Jose M.M.

    1993-01-01

    Axisymmetric spacetimes with a conformal symmetry are studied and it is shown that, if there is no further conformal symmetry, the axial Killing vector and the conformal Killing vector must commute. As a direct consequence, in conformally stationary and axisymmetric spacetimes, no restriction is made by assuming that the axial symmetry and the conformal timelike symmetry commute. Furthermore, we prove that in axisymmetric spacetimes with another symmetry (such as stationary and axisymmetric or cylindrically symmetric spacetimes) and a conformal symmetry, the commutator of the axial Killing vector with the two others mush vanish or else the symmetry is larger than that originally considered. The results are completely general and do not depend on Einstein's equations or any particular matter content.

  17. Axial flow positive displacement worm gas generator

    Science.gov (United States)

    Murrow, Kurt David (Inventor); Giffin, Rollin George (Inventor); Fakunle, Oladapo (Inventor)

    2010-01-01

    An axial flow positive displacement engine has an inlet axially spaced apart and upstream from an outlet. Inner and outer bodies have offset inner and outer axes extend from the inlet to the outlet through first, second, and third sections of a core assembly in serial downstream flow relationship. At least one of the bodies is rotatable about its axis. The inner and outer bodies have intermeshed inner and outer helical blades wound about the inner and outer axes respectively. The inner and outer helical blades extend radially outwardly and inwardly respectively. The helical blades have first, second, and third twist slopes in the first, second, and third sections respectively. The first twist slopes are less than the second twist slopes and the third twist slopes are less than the second twist slopes. A combustor section extends axially downstream through at least a portion of the second section.

  18. Improving the lattice axial vector current

    Energy Technology Data Exchange (ETDEWEB)

    Horsley, R. [Edinburgh Univ. (United Kingdom). School of Physics and Astronomy; Nakamura, Y. [RIKEN Advanced Institute for Computational Science, Kobe (Japan); Perlt, H.; Schiller, A. [Leipzig Univ. (Germany). Inst. fuer Theoretische Physik; Rakow, P.E.L. [Liverpool Univ. (United Kingdom). Theoretical Physics Div.; Schierholz, G. [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Zanotti, J.M. [Adelaide Univ. (Australia). CSSM, Dept. of Physics

    2015-11-15

    For Wilson and clover fermions traditional formulations of the axial vector current do not respect the continuum Ward identity which relates the divergence of that current to the pseudoscalar density. Here we propose to use a point-split or one-link axial vector current whose divergence exactly satisfies a lattice Ward identity, involving the pseudoscalar density and a number of irrelevant operators. We check in one-loop lattice perturbation theory with SLiNC fermion and gauge plaquette action that this is indeed the case including order O(a) effects. Including these operators the axial Ward identity remains renormalisation invariant. First preliminary results of a nonperturbative check of the Ward identity are also presented.

  19. Organocatalytic atroposelective synthesis of axially chiral styrenes

    Science.gov (United States)

    Zheng, Sheng-Cai; Wu, San; Zhou, Qinghai; Chung, Lung Wa; Ye, Liu; Tan, Bin

    2017-05-01

    Axially chiral compounds are widespread in biologically active compounds and are useful chiral ligands or organocatalysts in asymmetric catalysis. It is well-known that styrenes are one of the most abundant and principal feedstocks and thus represent excellent prospective building blocks for chemical synthesis. Driven by the development of atroposelective synthesis of axially chiral styrene derivatives, we discovered herein the asymmetric organocatalytic approach via direct Michael addition reaction of substituted diones/ketone esters/malononitrile to alkynals. The axially chiral styrene compounds were produced with good chemical yields, enantioselectivities and almost complete E/Z-selectivities through a secondary amine-catalysed iminium activation strategy under mild conditions. Such structural motifs are important precursors for further transformations into biologically active compounds and synthetic useful intermediates and may have potential applications in asymmetric synthesis as olefin ligands or organocatalysts.

  20. Improving the lattice axial vector current

    CERN Document Server

    Horsley, R; Perlt, H; Rakow, P E L; Schierholz, G; Schiller, A; Zanotti, J M

    2015-01-01

    For Wilson and clover fermions traditional formulations of the axial vector current do not respect the continuum Ward identity which relates the divergence of that current to the pseudoscalar density. Here we propose to use a point-split or one-link axial vector current whose divergence exactly satisfies a lattice Ward identity, involving the pseudoscalar density and a number of irrelevant operators. We check in one-loop lattice perturbation theory with SLiNC fermion and gauge plaquette action that this is indeed the case including order $O(a)$ effects. Including these operators the axial Ward identity remains renormalisation invariant. First preliminary results of a nonperturbative check of the Ward identity are also presented.

  1. Influence of the flux axial form on the conversion rate and duration of cycle between recharging for ThPu and U{sub nat} fuels in CANDU reactors; Influence de la forme axiale du flux sur le taux de conversion et la duree du cycle entre rechargements pour du combustible ThPu et U{sub nat} dans les reacteurs CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Chambon, Richard [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier / CNRS-IN2P3, 53 Avenue des Martyrs, F-38026 Grenoble (France)

    2007-01-15

    To face the increasing world power demand the world nuclear sector must be continuously updated and developed as well. Thus reactors of new types are introduced and advanced fuel cycles are proposed. The technological and economic feasibility and the transition of the present power park to a renewed park require thorough studies and scenarios, which are highly dependent on the reactor performances. The conversion rate and cycle span between recharging are important parameters in the scenarios studies. In this frame, we have studied the utilisation of thorium in the CANDU type reactors and particularly the influence of axial form of the flux, i.e. of the recharging mode, on the conversion rate and duration of the cycle between recharging. The results show that up to a first approximation the axial form of the flux resulting from the neutron transport calculations for assessing the conversion rate is not necessary to be taken into account. However the time span between recharging differs up to several percents if the axial form of the flux is taken into consideration in transport calculations. Thus if the burnup or the recharging frequency are parameters which influence significantly the deployment scenarios of a nuclear park an approach more refined than a simple transport evolution in a typical cell/assembly is recommended. Finally, the results of this study are not more general than for the assumed conditions but they give a thorough calculation method valid for any recharging/fuel combination in a CANDU type reactor.

  2. Axial mixing and mass transfer characteristics of pulsed extraction column with discs and doughnuts

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The axial mixing is a key factor for design and scaling up of the pulsed extraction column which has strong influence on the mass transfer performances of the pulsed extraction column with discs and doughnuts. A steady-state concentration profile measurement was used to evaluate the mass transfer and axial mixing coefficients for the nitric acid/water/30%TRPO (in kerosene) system in the pulsed extraction column with the diameter of 38 mm on the condition of the TRPO-kerosene solution as the continuous phase and the flow ratio of 1: 1. Experimental results indicate that Ey evaluated by the experiments is in good agreement with that given by Burratti's correlation; the axial mixing is far smaller for the continuous phase than that for the dispersed phase. However the empirical correlation for Hox is only given based on the present data.

  3. Numerical simulations of leading-edge vortex core axial velocity for flow over delta wings

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    Numerical simulations have been performed to investigate the characteristics of leading-edge vortex core axial velocity over two delta wings with leading edge swept angles Λ =50°and 76°, respectively. It is obtained that Reynolds number has the most important effect on the axial velocity of the primary leading-edge vortex core. At Reynolds numbers larger than 105, the jet-like flow of the vortex core is the most common type for both the large and the moderate swept delta wings. While if Reynolds number decreases to 103―104, the core axial velocity distributions for these two delta wings present the wake-like profile for all angles of attack considered in the present investigation.

  4. Axial loaded MRI of the lumbar spine

    Energy Technology Data Exchange (ETDEWEB)

    Saifuddin, A. E-mail: asaifuddin@aol.com; Blease, S.; MacSweeney, E

    2003-09-01

    Magnetic resonance imaging is established as the technique of choice for assessment of degenerative disorders of the lumbar spine. However, it is routinely performed with the patient supine and the hips and knees flexed. The absence of axial loading and lumbar extension results in a maximization of spinal canal dimensions, which may in some cases, result in failure to demonstrate nerve root compression. Attempts have been made to image the lumbar spine in a more physiological state, either by imaging with flexion-extension, in the erect position or by using axial loading. This article reviews the literature relating to the above techniques.

  5. Water Ingestion Into Axial Flow Compressors

    Science.gov (United States)

    1976-08-01

    AFAPL-TR-76-77 WATER INGESTION INTO AXIAL FLOW COMPRESSORS PURDUE UNIVERSITY SCHOOL OF AERONAUTICS AND ASTRONA UTICS S WEST LAFAYETTE, INDIANA 47907...CIPIENT’S CATALOG NUMBER TITL _07" 0 EREO Final i-7 0 Water Ingestion Into Axial Flow Compressorse 1 Auq 75 -: 31 Au0 a6 114o’ H-WPAFB-T-76-l:P ."CO TACT...necessary and Idenify by block number) Water ingestion , turbomachinery, and jet engines. 20 ABSTRACT (Contlinue on tov.ras side Hi necessary and Identify

  6. «FLARES» IN AXIAL SPONDYLOARTHRITIS

    Directory of Open Access Journals (Sweden)

    Sh. F. Erdes

    2016-01-01

    Full Text Available The clear definition of the concept of «flare in axial spondyloarthritis» is of paramount importance for clinical trials and routine practice in particular. It will be able to unify the characteristics of outcomes over a particular period of time on the one hand and to standardize therapeutic approaches on the other. On 4 February 2016, the journal Annals of Rheumatic Diseases published the on-line paper «Preliminary definitions of 'flare' in axial spondyloarthritis, based on pain, BASDAI and ASDAS-CRP: an ASAS initiative» by L. Gossec et al., which was devoted to this topic.

  7. Optimization of Axial Intensity Point Spread Function

    Institute of Scientific and Technical Information of China (English)

    WANG Haifeng; GAN Fuxi; CHEN Zhongyu

    2001-01-01

    It is known that for the converged laser beam, the axial intensity distribution corresponds to a Gaussian curve, that is, the intensity on the focal plane is the peak intensity. When it defocuses, the intensity would decrease rapidly. In optical data storage, for instance, we expect the intensity within a certain distance to be almost equal. In this paper, we propose to use a pure phase superresolution apodizer to optimize the axial intensity distribution of the converged laser beam and at the same time improve the resolution. The intensity point spread function remains almost identical in a wide range within the focal depth.

  8. Axial Nucleon form factors from lattice QCD

    CERN Document Server

    Alexandrou, C; Carbonell, J; Constantinou, M; Harraud, P A; Guichon, P; Jansen, K; Korzec, T; Papinutto, M

    2010-01-01

    We present results on the nucleon axial form factors within lattice QCD using two flavors of degenerate twisted mass fermions. Volume effects are examined using simulations at two volumes of spatial length $L=2.1$ fm and $L=2.8$ fm. Cut-off effects are investigated using three different values of the lattice spacings, namely $a=0.089$ fm, $a=0.070$ fm and $a=0.056$ fm. The nucleon axial charge is obtained in the continuum limit and chirally extrapolated to the physical pion mass enabling comparison with experiment.

  9. Axial Vircator for Electronic Warfare Applications

    Directory of Open Access Journals (Sweden)

    L. Drazan

    2009-12-01

    Full Text Available This paper deals with a high power microwave generator with virtual cathode – vircator in axial release for electronic warfare applications. The classification of directed energy weapons microwave (DEWM is introduced together with basic block diagrams of a particular class of DEWM. In the paper, methods for designing vircator pulsed power supply, axial vircator structure, measurement methods and experimental results are presented. The vircator in electromagnetic ammunition is powered by magneto-cumulative generator and in weapons for defense of objects (WDO, it is powered by Marx generator. The possible applications of a vircator in the DEWM area are discussed.

  10. Scalar Resonances in Axially Symmetric Spacetimes

    CERN Document Server

    Ranea-Sandoval, Ignacio F

    2015-01-01

    We study properties of resonant solutions to the scalar wave equation in several axially symmetric spacetimes. We prove that non-axial resonant modes do not exist neither in the Lanczos dust cylinder, the $(2+1)$ extreme BTZ spacetime nor in a class of simple rotating wormhole solutions. Moreover, we find unstable solutions to the wave equation in the Lanczos dust cylinder and in the $r^2 <0$ region of the extreme $(2+1)$ BTZ spacetime, two solutions that possess closed timelike curves. Similarities with previous results obtained for the Kerr spacetime are explored.

  11. Spiral and Taylor vortex fronts and pulses in axial through flow.

    Science.gov (United States)

    Pinter, A; Lücke, M; Hoffmann, Ch

    2003-02-01

    The influence of an axial through flow on the spatiotemporal growth behavior of different vortex structures in the Taylor-Couette system with radius ratio eta=0.5 is determined. The Navier-Stokes equations (NSE) linearized around the basic Couette-Poiseuille flow are solved numerically with a shooting method in a wide range of through flow strengths Re and different rates of co-rotating and counter-rotating cylinders for toroidally closed vortices with azimuthal wave number m=0 and for spiral vortex flow with m=+/-1. For each of these three different vortex varieties we have investigated (i) axially extended vortex structures, (ii) axially localized vortex pulses, and (iii) vortex fronts. The complex dispersion relations of the linearized NSE for vortex modes with the three different m are evaluated for real axial wave numbers for (i) and over the plane of complex axial wave numbers for (ii) and (iii). We have also determined the Ginzburg-Landau amplitude equation (GLE) approximation in order to analyze its predictions for the vortex structures (ii) and (iii). Critical bifurcation thresholds for extended vortex structures are evaluated. The boundaries between absolute and convective instability of the basic state for vortex pulses are determined with a saddle-point analysis of the dispersion relations. Fit parameters for power-law expansions of the boundaries up to Re4 are listed in two tables. Finally, the linearly selected front behavior of growing vortex structures is investigated using saddle-point analyses of the dispersion relations of NSE and GLE. For the two front intensity profiles (increasing in positive or negative axial direction) we have determined front velocities, axial growth rates, and the wave numbers and frequencies of the unfolding vortex patterns with azimuthal wave numbers m=0,+/-1, respectively.

  12. Axial and radial velocities in the creeping flow in a pipe

    Directory of Open Access Journals (Sweden)

    Zuykov Andrey L'vovich

    2014-05-01

    Full Text Available The article is devoted to analytical study of transformation fields of axial and radial velocities in uneven steady creeping flow of a Newtonian fluid in the initial portion of the cylindrical channel. It is shown that the velocity field of the flow is two-dimensional and determined by the stream function. The article is a continuation of a series of papers, where normalized analytic functions of radial axial distributions in uneven steady creeping flow in a cylindrical tube with azimuthal vorticity and stream function were obtained. There is Poiseuille profile for the axial velocity in the uniform motion of a fluid at an infinite distance from the entrance of the pipe (at x = ∞, here taken equal to zero radial velocity. There is uniform distribution of the axial velocity in the cross section at the tube inlet at x = 0, at which the axial velocity is constant along the current radius. Due to the axial symmetry of the flow on the axis of the pipe (at r = 0, the radial velocities and the partial derivative of the axial velocity along the radius, corresponding to the condition of the soft function extremum, are equal to zero. The authors stated vanishing of the velocity of the fluid on the walls of the pipe (at r = R , where R - radius of the tube due to its viscous sticking and tightness of the walls. The condition of conservation of volume flow along the tube was also accepted. All the solutions are obtained in the form of the Fourier - Bessel. It is shown that the hydraulic losses at uniform creeping flow of a Newtonian fluid correspond to Poiseuille - Hagen formula.

  13. Axial T2* mapping in intervertebral discs: a new technique for assessment of intervertebral disc degeneration

    Energy Technology Data Exchange (ETDEWEB)

    Hoppe, Sven; Quirbach, Sebastian; Krause, Fabian G.; Benneker, Lorin M. [Inselspital, Berne University Hospital, Department of Orthopaedic Surgery, Berne (Switzerland); Mamisch, Tallal C. [Inselspital, Berne University Hospital, Department of Radiology, Berne (Switzerland); Werlen, Stefan [Clinic Sonnenhof, Department of Radiology, Berne (Switzerland)

    2012-09-15

    To demonstrate the potential benefits of biochemical axial T2* mapping of intervertebral discs (IVDs) regarding the detection and grading of early stages of degenerative disc disease using 1.5-Tesla magnetic resonance imaging (MRI) in a clinical setting. Ninety-three patients suffering from lumbar spine problems were examined using standard MRI protocols including an axial T2* mapping protocol. All discs were classified morphologically and grouped as ''healthy'' or ''abnormal''. Differences between groups were analysed regarding to the specific T2* pattern at different regions of interest (ROIs). Healthy intervertebral discs revealed a distinct cross-sectional T2* value profile: T2* values were significantly lower in the annulus fibrosus compared with the nucleus pulposus (P = 0.01). In abnormal IVDs, T2* values were significantly lower, especially towards the centre of the disc representing the expected decreased water content of the nucleus (P = 0.01). In herniated discs, ROIs within the nucleus pulposus and ROIs covering the annulus fibrosus showed decreased T2* values. Axial T2* mapping is effective to detect early stages of degenerative disc disease. There is a potential benefit of axial T2* mapping as a diagnostic tool, allowing the quantitative assessment of intervertebral disc degeneration. circle Axial T2* mapping effective in detecting early degenerative disc disease. (orig.)

  14. LDA measurements on the turbulent flow characteristics of a small-sized axial fan

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jang Kweon [Kunsan National Univ., Kunsan (Korea, Republic of)

    2001-07-01

    The operating point of a small-sized axial fan for refrigerator is strongly dependent upon the system resistance. Therefore, the turbulent flow characteristics around a small-sized axial fan may change significantly according to the operating point. This study represents three-dimensional turbulent flow characteristics around a small-sized axial fan measured at the four operating points such as {phi}=0.1, 0.18, 0.25 and 0.32 by using fiber-optic type LDA system. This LDA system is composed of a 5 W Argon-ion laser, two optics in back-scatter mode, three BSA's, a PC, and a three-dimensional automatic traversing system. A kind of paraffin fluid is utilized for supplying particles by means of fog generator. Mean velocity profiles downstream of a small--sized axial fan along the radial distance show that both the streamwise and the tangential components exist predominantly in downstream except {phi}=0.1 and have a maximum value at the radial distance ratio of about 0.8, but the radial component, which its velocity is relatively small, is acting role that only turns flow direction to the outside or the central part of axial fan. Moreover, all of the velocity components downstream at {phi}=0.1 show much smaller than those upstream due to the static pressure rise at the low-flowrate region.

  15. 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

  16. Resolving optical illumination distributions along an axially symmetric photodetecting fiber.

    Science.gov (United States)

    Sorin, Fabien; Lestoquoy, Guillaume; Danto, Sylvain; Joannopoulos, John D; Fink, Yoel

    2010-11-08

    Photodetecting fibers of arbitrary length with internal metal, semiconductor and insulator domains have recently been demonstrated. These semiconductor devices exhibit a continuous translational symmetry which presents challenges to the extraction of spatially resolved information. Here, we overcome this seemingly fundamental limitation and achieve the detection and spatial localization of a single incident optical beam at sub-centimeter resolution, along a one-meter fiber section. Using an approach that breaks the axial symmetry through the constuction of a convex electrical potential along the fiber axis, we demonstrate the full reconstruction of an arbitrary rectangular optical wave profile. Finally, the localization of up to three points of illumination simultaneously incident on a photodetecting fiber is achieved.

  17. An Axial Dispersion Model for Evaporating Bubble Column Reactor

    Institute of Scientific and Technical Information of China (English)

    谢刚; 李希

    2004-01-01

    Evaporating bubble column reactor (EBCR) is a kind of aerated reactor in which the reaction heat is removed by the evaporation of volatile reaction mixture. In this paper, a mathematical model that accounts for the gas-liquid exothermic reaction and axial dispersions of both gas and liquid phase is employed to study the performance of EBCR for the process of p-xylene(PX) oxidation. The computational results show that there are remarkable concentration and temperature gradients in EBCR for high ratio of height to diameter (H/DT). The temperature is lower at the bottom of column and higher at the top, due to rapid evaporation induced by the feed gas near the bottom. The concentration profiles in the gas phase are more nonuniform than those (except PX) in the liquid phase, which causes more solvent burning consumption at high H/DT ratio. For p-xylene oxidation, theo ptimal H/DT is around 5.

  18. Excitation modes in non-axial nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Leviatan, A.; Ginnochio, J.N.

    1990-01-01

    Excitation modes of non-axial quadrupole shapes are investigated in the framework of interacting boson models. Both {gamma}-unstable and {gamma}-rigid nuclear shapes are considered for systems with one type of boson as well as with proton-neutron bosons. 6 refs.

  19. Cystic lesions accompanying extra-axial tumours

    NARCIS (Netherlands)

    Lohle, PNM; Wurzer, HAL; Seelen, PJ; Kingma, LM; Go, KG

    1999-01-01

    We examined the mechanism of cyst formation in extra-axial tumours in the central nervous system (CNS). Cyst fluid, cerebrospinal fluid (CSF) and blood plasma were analysed in eight patients with nine peritumoral cysts: four with meningiomas, two with intracranial and two spinal intradural schwannom

  20. Knowledge Based Design of Axial Flow Compressor

    Directory of Open Access Journals (Sweden)

    Dinesh kumar.R

    2015-05-01

    Full Text Available In the aerospace industry with highly competitive market the time to design and delivery is shortening every day. Pressure on delivering robust product with cost economy is in demand in each development. Even though technology is older, it is new for each customer requirement and highly non-liner to fit one in another place. Gas turbine is considered one of a complex design in the aircraft system. It involves experts to be grouped with designers of various segments to arrive the best output. The time is crucial to achieve a best design and it needs knowledge automation incorporated with CAD/CAE tools. In the present work an innovative idea in the form of Knowledge Based Engineering for axial compressor is proposed, this includes the fundamental design of axial compressor integrated with artificial intelligence in the form of knowledge capturing and programmed with high level language (Visual Basis.Net and embedded into CATIA v5. This KBE frame work eases out the design and modeling of axial compressor design and produces 3D modeling for further flow simulation with fluid dynamic in Ansys-Fluent. Most of the aerospace components are developed through simulation driven product development and in this case it is established for axial compressor.

  1. Active axial stress in mouse aorta.

    Science.gov (United States)

    Agianniotis, A; Rachev, A; Stergiopulos, N

    2012-07-26

    The study verifies the development of active axial stress in the wall of mouse aorta over a range of physiological loads when the smooth muscle cells are stimulated to contract. The results obtained show that the active axial stress is virtually independent of the magnitude of pressure, but depends predominately on the longitudinal stretch ratio. The dependence is non-monotonic and is similar to the active stress-stretch dependence in the circumferential direction reported in the literature. The expression for the active axial stress fitted to the experimental data shows that the maximum active stress is developed at longitudinal stretch ratio 1.81, and 1.56 is the longitudinal stretch ratio below which the stimulation does not generate active stress. The study shows that the magnitude of active axial stress is smaller than the active circumferential stress. There is need for more experimental investigations on the active response of different types of arteries from different species and pathological conditions. The results of these studies can promote building of refined constrictive models in vascular rheology. Copyright © 2012 Elsevier Ltd. All rights reserved.

  2. Investigations on Experimental Impellers for Axial Blowers

    Science.gov (United States)

    Encke, W.

    1947-01-01

    A selection of measurements obtained on experimental impellers for axial blowers will be reported. In addition to characteristic curves plotted for low and for high peripheral velocities, proportions and blade sections for six different blower models and remarks on the design of blowers will be presented.

  3. Wave propagation in axially moving periodic strings

    DEFF Research Database (Denmark)

    Sorokin, Vladislav S.; Thomsen, Jon Juel

    2017-01-01

    The paper deals with analytically studying transverse waves propagation in an axially moving string with periodically modulated cross section. The structure effectively models various relevant technological systems, e.g. belts, thread lines, band saws, etc., and, in particular, roller chain drive...

  4. The Kenya rift axial gravity high: a re-interpretation

    Science.gov (United States)

    Swain, C. J.

    1992-03-01

    Since KRISP 85 did not provide overwhelming evidence for the massive intrusion that was originally suggested to explain the axial gravity high yet did provide a velocity section for the upper crust along the axis of the Kenya Rift, it is appropriate to use this section to control a re-interpretation of the gravity anomalies. A 2 {1}/{2} D inversion procedure has been used to model a number of isostatic anomaly profiles between Lake Baringo and Suswa. There are too many unknowns and gravity station coverage is too sparse for the results to be unique. Nevertheless, certain conclusions can be drawn. One of those is that some relatively dense material exists within the basement all along the Rift axis, since the axial isostatic anomalies are positive even though they occur where there are several thousand metres of Cenozoic volcanics of relatively low density (inferred from their seismic velocity of 3.7-5.1 km/s). The dense material is envisaged as a zone of dyke injection and assigned a density of 2.75-2.76 g/cm 3 (corresponding to its 6.05 km/s velocity) compared to a normal basement density of 2.70 g/cm 3. It is assumed to extend down to 22 km—the top of the 7.1 km/s layer. The KRISP 85 line passed just east of Menengai, where the basement velocity increases to about 6.6 km/s over a distance of about 20 km. On an east-west gravity profile through Menengai there is a gravity high corresponding to this velocity increase which has been modelled as a basic intrusion (density 2.93 g/cm 3) underlying the caldera.

  5. Axial low back pain: one painful area--many perceptions and mechanisms.

    Directory of Open Access Journals (Sweden)

    Matti Förster

    Full Text Available Axial low back pain can be considered as a syndrome with both nociceptive and neuropathic pain components (mixed-pain. Especially neuropathic pain comprises a therapeutic challenge in practical experience and may explain why pharmacotherapy in back pain is often disappointing for both the patient and the therapist. This survey uses epidemiological and clinical data on the symptomatology of 1083 patients with axial low back pain from a cross sectional survey (painDETECT. Objectives were (1 to estimate whether neuropathic pain contributes to axial low back pain and if so to what extent. (2 To detect subgroups of patients with typical sensory symptom profiles and to analyse their demographic data and co-morbidities. (3 To compare patients with and without prior intervertebral disc surgery (IVD. Neuropathic pain components could be detected in 12% of the entire cohort. Cluster analyses of these patients revealed five distinct subgroups of patients showing a characteristic sensory profile, i.e. a typical constellation and combination of symptoms. All subgroups occurred in relevant numbers and some showed distinct neuropathic characteristics while others showed nociceptive features. Post-IVD-surgery patients showed a tendency to score more "neuropathic" than patients without surgery (not statistically significant. Axial low back pain has a high prevalence of co-morbidities with implication on therapeutic aspects. From these data it can be concluded that sensory profiles based on descriptor severity may serve as a better predictor for therapy assessment than pain intensity or sole diagnosis alone. Standardized phenotyping of pain symptoms with easy tools may help to develop an individualized therapy leading to a higher success rate in pharmacotherapy of axial low back pain.

  6. Might axial myofascial properties and biomechanical mechanisms be relevant to ankylosing spondylitis and axial spondyloarthritis?

    OpenAIRE

    Masi, Alfonse T.

    2014-01-01

    Ankylosing spondylitis and axial spondyloarthropathy have characteristic age- and sex-specific onset patterns, typical entheseal lesions, and marked heritability, but the integrative mechanisms causing the pathophysiological and structural alterations remain largely undefined. Myofascial tissues are integrated in the body into webs and networks which permit transmission of passive and active tensional forces that provide stabilizing support and help to control movements. Axial myofascial hype...

  7. Might axial myofascial properties and biomechanical mechanisms be relevant to ankylosing spondylitis and axial spondyloarthritis?

    OpenAIRE

    Masi, Alfonse T.

    2014-01-01

    Ankylosing spondylitis and axial spondyloarthropathy have characteristic age- and sex-specific onset patterns, typical entheseal lesions, and marked heritability, but the integrative mechanisms causing the pathophysiological and structural alterations remain largely undefined. Myofascial tissues are integrated in the body into webs and networks which permit transmission of passive and active tensional forces that provide stabilizing support and help to control movements. Axial myofascial hype...

  8. Determination of burnup grade of fuel plates by gamma spectrometry; Determinacao do grau de queima em elementos combustiveis tipo placa por meio de espectrometria gama

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis A.A.; Zeituni, Carlos A.; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1999-11-01

    This work describes absolute burnup measurements on spent MTR fuel elements by means of non-destructive gamma-ray spectroscopy which correlates activities of radioactive fission products with the fissioned mass of {sup 235} U. Experiments based on such method were performed at the storage pool area of the IEA-R1 research reactor. The obtained results were compared with calculational ones based on neutronics. (author) 7 refs., 6 figs., 1 tab.; e-mail: laaterre at net.ipen.br

  9. 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.

  10. 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)

  11. 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)

  12. A model of axial heterostructure formation in III-V semiconductor nanowires

    Science.gov (United States)

    Dubrovskii, V. G.

    2016-03-01

    A kinetic model of the formation of axial heterostructures in nanocrystalline wires (nanowires, NWs) of III-V semiconductor compounds growing according to the vapor-liquid-solid (VLS) mechanism is proposed. A general system of nonstationary equations for effective fluxes of two elements of the same group (e.g., group III) is formulated that allows the composition profile of a heterostructure to be calculated as a function of the coordinate and epitaxial growth conditions, including the flux of a group V element. Characteristic times of the composition relaxation, which determine the sharpness of the heteroboundary (heterointerface), are determined in the linear approximation. A temporal interruption (arrest) of fluxes during the switching of elements for a period exceeding these relaxation times must increase sharpness of the heteroboundary. Model calculations of the composition profile in a double GaAs/InAs/GaAs axial heterostructure have been performed for various NW radii.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-15

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

  14. 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)

  15. 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).

  16. Piping inspection carriage having axially displaceable sensor

    Science.gov (United States)

    Zollinger, William T.; Treanor, Richard C.

    1994-01-01

    A pipe inspection instrument carriage for use with a pipe crawler for performing internal inspections of piping surfaces. The carriage has a front leg assembly, a rear leg assembly and a central support connecting the two assemblies and for mounting an instrument arm having inspection instruments. The instrument arm has a y-arm mounted distally thereon for axially aligning the inspection instrumentation and a mounting block, a linear actuator and axial movement arm for extending the inspection instruments radially outward to operably position the inspection instruments on the piping interior. Also, the carriage has a rotation motor and gear assembly for rotating the central support and the front leg assembly with respect to the rear leg assembly so that the inspection instruments azimuthally scan the piping interior. The instrument carriage allows performance of all piping inspection operations with a minimum of moving parts, thus decreasing the likelihood of performance failure.

  17. Direct optical nanoscopy with axially localized detection

    CERN Document Server

    Bourg, N; Dupuis, G; Barroca, T; Bon, P; Lécart, S; Fort, E; Lévêque-Fort, S

    2014-01-01

    Evanescent light excitation is widely used in super-resolution fluorescence microscopy to confine light and reduce background noise. Herein we propose a method of exploiting evanescent light in the context of emission. When a fluorophore is located in close proximity to a medium with a higher refractive index, its near-field component is converted into light that propagates beyond the critical angle. This so-called Supercritical Angle Fluorescence (SAF) can be captured using a hig-NA objective and used to determine the axial position of the fluorophore with nanometer precision. We introduce a new technique for 3D nanoscopy that combines direct STochastic Optical Reconstruction Microscopy (dSTORM) imaging with dedicated detection of SAF emission. We demonstrate that our approach of a Direct Optical Nanoscopy with Axially Localized Detection (DONALD) yields a typical isotropic 3D localization precision of 20 nm.

  18. Axial flow positive displacement worm compressor

    Science.gov (United States)

    Murrow, Kurt David (Inventor); Giffin, Rollin George (Inventor); Fakunle, Oladapo (Inventor)

    2010-01-01

    An axial flow positive displacement compressor has an inlet axially spaced apart and upstream from an outlet. Inner and outer bodies have offset inner and outer axes extend from the inlet to the outlet through first and second sections of a compressor assembly in serial downstream flow relationship. At least one of the bodies is rotatable about its axis. The inner and outer bodies have intermeshed inner and outer helical blades wound about the inner and outer axes respectively. The inner and outer helical blades extend radially outwardly and inwardly respectively. The helical blades have first and second twist slopes in the first and second sections respectively. The first twist slopes are less than the second twist slopes. An engine including the compressor has in downstream serial flow relationship from the compressor a combustor and a high pressure turbine drivingly connected to the compressor by a high pressure shaft.

  19. Matrix calculus for axially symmetric polarized beam.

    Science.gov (United States)

    Matsuo, Shigeki

    2011-06-20

    The Jones calculus is a well known method for analyzing the polarization of a fully polarized beam. It deals with a beam having spatially homogeneous polarization. In recent years, axially symmetric polarized beams, where the polarization is not homogeneous in its cross section, have attracted great interest. In the present article, we show the formula for the rotation of beams and optical elements on the angularly variant term-added Jones calculus, which is required for analyzing axially symmetric beams. In addition, we introduce an extension of the Jones calculus: use of the polar coordinate basis. With this calculus, the representation of some angularly variant beams and optical elements are simplified and become intuitive. We show definitions, examples, and conversion formulas between different notations.

  20. Consistent formulation of the spacelike axial gauge

    Energy Technology Data Exchange (ETDEWEB)

    Burnel, A.; Van der Rest-Jaspers, M.

    1983-12-15

    The usual formulation of the spacelike axial gauge is afflicted with the difficulty that the metric is indefinite while no ghost is involved. We solve this difficulty by introducing a ghost whose elimination is such that the metric becomes positive for physical states. The technique consists in the replacement of the gauge condition nxA = 0 by the weaker one partial/sub 0/nxAroughly-equal0.

  1. Transonic Axial Splittered Rotor Tandem Stator Stage

    Science.gov (United States)

    2016-12-01

    compressor rotor was designed incorporating a splitter vane between the principal blades . Historical experiments conducted by Dr. Arthur J...conventional rotor design . The stage is composed of the rotor and stator. The flow of the air passing through the rotor is turned, and the flow is required...derived results achieved the best blade geometry for design continuation. The best circumferential and axial placement for the splitter blade was

  2. Multimode interaction in axially excited cylindrical shells

    OpenAIRE

    2014-01-01

    Cylindrical shells exhibit a dense frequency spectrum, especially near the lowest frequency range. In addition, due to the circumferential symmetry, frequencies occur in pairs. So, in the vicinity of the lowest natural frequencies, several equal or nearly equal frequencies may occur, leading to a complex dynamic behavior. So, the aim of the present work is to investigate the dynamic behavior and stability of cylindrical shells under axial forcing with multiple equal or nearly equal natural fr...

  3. Axial flux permanent magnet brushless machines

    CERN Document Server

    Gieras, Jacek F; Kamper, Maarten J

    2008-01-01

    Axial Flux Permanent Magnet (AFPM) brushless machines are modern electrical machines with a lot of advantages over their conventional counterparts. They are being increasingly used in consumer electronics, public life, instrumentation and automation system, clinical engineering, industrial electromechanical drives, automobile manufacturing industry, electric and hybrid electric vehicles, marine vessels and toys. They are also used in more electric aircrafts and many other applications on larger scale. New applications have also emerged in distributed generation systems (wind turbine generators

  4. Enhancement of Optical Coherence Tomography Axial Resolution by Spectral Shaping

    Institute of Scientific and Technical Information of China (English)

    孙汕; 郭继华; 高湔松; 薛平

    2002-01-01

    We propose a new method of changing the spectrum shape to improve the axial resolution of optical coherencetomography (OCT). Theoretical analysis shows that certain spectral shaping can shorten the coherence length.Comparisons of the simulation and experimental measurements of spectral shape and axial resolution of OCTare given, showing that the axial resolution of OCT is enhanced by a factor of 1.4.

  5. Direct optical nanoscopy with axially localized detection

    Science.gov (United States)

    Bourg, N.; Mayet, C.; Dupuis, G.; Barroca, T.; Bon, P.; Lécart, S.; Fort, E.; Lévêque-Fort, S.

    2015-09-01

    Evanescent light excitation is widely used in super-resolution fluorescence microscopy to confine light and reduce background noise. Here, we propose a method of exploiting evanescent light in the context of emission. When a fluorophore is located in close proximity to a medium with a higher refractive index, its near-field component is converted into light that propagates beyond the critical angle. This so-called supercritical-angle fluorescence can be captured using a high-numerical-aperture objective and used to determine the axial position of the fluorophore with nanometre precision. We introduce a new technique for three-dimensional nanoscopy that combines direct stochastic optical reconstruction microscopy (dSTORM) with dedicated detection of supercritical-angle fluorescence emission. We demonstrate that our approach of direct optical nanoscopy with axially localized detection (DONALD) typically yields an isotropic three-dimensional localization precision of 20 nm within an axial range of ∼150 nm above the coverslip.

  6. Axial gravity, massless fermions and trace anomalies

    Science.gov (United States)

    Bonora, L.; Cvitan, M.; Prester, P. Dominis; Pereira, A. Duarte; Giaccari, S.; Štemberga, T.

    2017-08-01

    This article deals with two main topics. One is odd parity trace anomalies in Weyl fermion theories in a 4d curved background, the second is the introduction of axial gravity. The motivation for reconsidering the former is to clarify the theoretical background underlying the approach and complete the calculation of the anomaly. The reference is in particular to the difference between Weyl and massless Majorana fermions and to the possible contributions from tadpole and seagull terms in the Feynman diagram approach. A first, basic, result of this paper is that a more thorough treatment, taking account of such additional terms and using dimensional regularization, confirms the earlier result. The introduction of an axial symmetric tensor besides the usual gravitational metric is instrumental to a different derivation of the same result using Dirac fermions, which are coupled not only to the usual metric but also to the additional axial tensor. The action of Majorana and Weyl fermions can be obtained in two different limits of such a general configuration. The results obtained in this way confirm the previously obtained ones.

  7. Golimumab for the treatment of axial spondyloarthritis.

    Science.gov (United States)

    Gelfer, Gita; Perry, Lisa; Deodhar, Atul

    2016-01-01

    Axial spondyloarthritis (axSpA) is a chronic, immune-mediated inflammatory disease of the axial skeleton that includes ankylosing spondylitis (AS) and non-radiographic axial spondyloarthritis (nr-axSpA). Patients with AS experience chronic pain due to sacroiliac joint and spinal inflammation, and may develop spinal ankylosing with syndesmophyte formation. Tumor necrosis factor α inhibitors (TNFi) have shown promise in the management of AS and axSpA by targeting the underlying inflammatory process, and providing symptomatic relief. Whether they alter the progression of the disease is uncertain. Golimumab is a fully human IgG1 monoclonal antibody that targets and downregulates the pro-inflammatory cytokine TNF-α. The use of golimumab has been shown to reduce the signs and symptoms of axSpA as well as improve patient function and quality reported outcomes. This review focuses on the biological rationale and the results of clinical trials with golimumab for the treatment of axSpA.

  8. Dynamic control of knee axial deformities

    Directory of Open Access Journals (Sweden)

    E. E. Malyshev

    2013-01-01

    Full Text Available The authors have evaluated the clinical examination of the patients with axial malalignments in the knee by the original method and device which was named varovalgometer. The measurements were conducted by tension of the cord through the spina iliaca anterior superior and the middle of the lower pole of patella. The deviation of the center of the ankle estimated by metal ruler which was positioned perpendicular to the lower leg axis on the level of the ankle joint line. The results of comparison of our method and computer navigation in 53 patients during the TKA show no statistically significant varieties but they differ by average 5° of valgus in clinical examination in comparison with mechanical axis which was identified by computer navigation. The dynamic control of axial malalignment can be used in clinical practice for estimation of the results of treatment of pathology with axial deformities in the knee; for the control of reduction and secondary displacement of the fractures around the knee; for assessment of instability; in planning of correctional osteotomies and intraoperative control of deformity correction; for estimation of Q angle in subluxation and recurrent dislocation of patella; in planning of TKA; during the growth of child it allows to assess the progression of deformity.

  9. Normative segment-specific axial and coronal angulation corridors of subaxial cervical column in axial rotation.

    Science.gov (United States)

    Yoganandan, Narayan; Stemper, Brian D; Pintar, Frank A; Baisden, Jamie L; Shender, Barry S; Paskoff, Glenn

    2008-03-01

    In contrast to clinical studies wherein loading magnitudes are indeterminate, experiments permit controlled and quantifiable moment applications, record kinematics in multiple planes, and allow derivation of moment-angulation corridors. Axial and coronal moment-angulation corridors were determined at every level of the subaxial cervical spine, expressed as logarithmic functions, and level-specificity of range of motion and neutral zones were evaluated. segmental primary axial and coupled coronal motions do not vary by level. Although it is known that cervical spine responses are coupled, segment-specific corridors of axial and coronal kinematics under axial twisting moments from healthy normal spines are not reported. Ten human cadaver columns (23-44 years, mean: 34 +/- 6.8) were fixed at the ends and targets were inserted to each vertebra to obtain kinematics in axial and coronal planes. The columns were subjected to pure axial twisting moments. Range of motion and neutral zone for primary-axial and coupled-coronal rotation components were determined at each spinal level. Data were analyzed using factorial analysis of variance. Moment-rotation angulations were expressed using logarithmic functions, and mean +/-1 standard deviation corridors were derived at each level for both components. Moment-angulations responses were nonlinear. Each segmental curve for both components was well represented by a logarithmic function (r2 > 0.95). Factorial analysis of variance indicated that the biomechanical metrics are spinal level-specific (P specific responses. The presentation of moment-angulation corridors for both metrics forms a dataset for the normal population. These segment-specific nonlinear corridors may help clinicians assess dysfunction or instability. These data will assist mathematical models of the spine in improved validation and lead to efficacious design of stabilizing systems.

  10. Might axial myofascial properties and biomechanical mechanisms be relevant to ankylosing spondylitis and axial spondyloarthritis?

    Science.gov (United States)

    Masi, Alfonse T

    2014-01-01

    Ankylosing spondylitis and axial spondyloarthropathy have characteristic age- and sex-specific onset patterns, typical entheseal lesions, and marked heritability, but the integrative mechanisms causing the pathophysiological and structural alterations remain largely undefined. Myofascial tissues are integrated in the body into webs and networks which permit transmission of passive and active tensional forces that provide stabilizing support and help to control movements. Axial myofascial hypertonicity was hypothesized as a potential excessive polymorphic trait which could contribute to chronic biomechanical overloading and exaggerated stresses at entheseal sites. Such a mechanism may help to integrate many of the characteristic host, pathological, and structural features of ankylosing spondylitis and axial spondyloarthritis. Biomechanical stress and strain were recently documented to correlate with peripheral entheseal inflammation and new bone formation in a murine model of spondyloarthritis. Ankylosing spondylitis has traditionally been classified by the modified New York criteria, which require the presence of definite radiographic sacroiliac joint lesions. New classification criteria for axial spondyloarthritis now include patients who do not fulfill the modified New York criteria. The male-to-female sex ratios clearly differed between the two patient categories - 2:1 or 3:1 in ankylosing spondylitis and 1:1 in non-radiographic axial spondyloarthritis - and this suggests a spectral concept of disease and, among females, milder structural alterations. Magnetic resonance imaging of active and chronic lesions in ankylosing spondylitis and axial spondyloarthritis reveals complex patterns, usually interpreted as inflammatory reactions, but shows similarities to acute degenerative disc disease, which attributed to edema formation following mechanical stresses and micro-damage. A basic question is whether mechanically induced microinjury and immunologically mediated

  11. Effects of Axial Non-uniform Tip Clearances on Aerodynamic Performance of a Transonic Axial Compressor

    Institute of Scientific and Technical Information of China (English)

    Hongwei MA; Baihe LI

    2008-01-01

    This paper presents a numerical investigation of effects of axial non-uniform tip clearances on the aerodynamic performance of a transonic axial compressor rotor (NASA Rotor 37). The three-dimensional steady flow field within the rotor passage was simulated with the datum tip clearance of 0.356 mm at the design wheel speed of 17188.7 rpm. The simulation results are well consistent with the measurement results, which verified the numeri-cal method. Then the three-dimensional steady flow field within the rotor passage was simulated respectively with different axial non-uniform tip clearances. The calculation results showed that optimal axial non-uniform tip clearances could improve the compressor performance, while the efficiency and the pressure ratio of the com-pressor were increased. The flow mechanism is that the axial non-uniform tip clearance can weaken the tip leak-age vortex, blow down low-energy fluids in boundary layers and reduce both flow blockage and tip loss.

  12. Intrinsic carpal ligaments on MR and multidetector CT arthrography: comparison of axial and axial oblique planes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ryan K.L.; Griffith, James F.; Ng, Alex W.H.; Law, Eric K.C. [The Chinese University of Hong Kong, Department of Imaging and Interventional Radiology, Prince Of Wales Hospital, Hong Kong (China); Tse, W.L.; Wong, Clara W.Y.; Ho, P.C. [The Chinese University of Hong Kong, Department of Orthopedics and Traumatology, Prince Of Wales Hospital, Hong Kong (China)

    2017-03-15

    To compare axial and oblique axial planes on MR arthrography (MRA) and multidetector CT arthrography (CTA) to evaluate dorsal and volar parts of scapholunate (SLIL) and lunotriquetral interosseous (LTIL) ligaments. Nine cadaveric wrists of five male subjects were studied. The visibility of dorsal and volar parts of the SLIL and LTIL was graded semi-quantitatively (good, intermediate, poor) on MRA and CTA. The presence of a ligament tear was determined on arthrosocopy and sensitivity, specificity and accuracy of tear detection were calculated. Oblique axial imaging was particularly useful for delineating dorsal and volar parts of the LTIL on MRA with overall 'good' visibility increased from 11 % to 78 %. The accuracy of MRA and CTA in revealing SLIL and LTIL tear was higher using the oblique axial plane. The overall accuracy for detecting SLIL tear on CTA improved from 94 % to 100 % and from 89 % to 94 % on MRA; the overall accuracy of detecting LTIL tear on CTA improved from 89 % to 100 % and from 72 % to 89 % on MRA Oblique axial imaging during CT and MR arthrography improves detection of tears in the dorsal and volar parts of both SLIL and LTIL. (orig.)

  13. Global axial-torsional dynamics during rotary drilling

    Science.gov (United States)

    Gupta, Sunit K.; Wahi, Pankaj

    2016-08-01

    We have studied the global dynamics of the bottom hole assembly (BHA) during rotary drilling with a lumped parameter axial-torsional model for the drill-string and a linear cutting force model. Our approach accounts for bit-bounce and stick-slip along with the regenerative effect and is independent of the drill-string and the bit-rock interaction model. Regenerative axial dynamics due to variable depth of cut is incorporated through a functional description of the cut surface profile instead of a delay differential equation with a state-dependent delay. The evolution of the cut surface is governed by a nonlinear partial differential equation (PDE) which is coupled with the ordinary differential equations (ODEs) governing the longitudinal and angular dynamics of the BHA. The boundary condition for the PDE captures multiple regeneration in the event of bit-bounce. Interruption in the torsional dynamics is included by considering separate evolution equations for the various states during the stick period. Finite-dimensional approximation for our coupled PDE-ODE model has been obtained and validated by comparing our results against existing results. Bifurcation analysis of our system reveals a supercritical Hopf bifurcation leading to periodic vibrations without bit-bounce and stick-slip which is followed by solutions involving bit-bounce or stick-slip depending on the operating parameters. Further inroads into the unstable regime leads to a variety of complex behavior including co-existence of periodic and chaotic solutions involving both bit-bounce and stick-slip.

  14. Co-axial multicusp source for low axial energy spread ion beam production

    CERN Document Server

    Lee, Y; Leung, K N; Vujic, J L; Williams, M D; Zahir, N

    1999-01-01

    A co-axial multicusp ion source has been designed and tested. This source uses a new magnetic filter configuration. This magnetic filter is efficient in modifying the plasma potential distribution which can reduce the axial energy spread of the extracted ion beam. Energy spreads as low as 0.6 eV have been obtained. The electron temperature in this source has also been found to be about 0.1 eV. Furthermore, the new source configuration is capable of adjusting the radial plasma potential distribution which can improve the transverse ion energy, which results in a low beam emittance. The co-axial source can be used for a number of different applications such as ion projection lithography and radioactive ion beam projects.

  15. Co-axial multicusp source for low axial energy spread ion beam production

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. E-mail: yylee@lbl.gov; Gough, R.A.; Leung, K.N.; Vujic, J.; Williams, M.D.; Zahir, N

    1999-09-01

    A co-axial multicusp ion source has been designed and tested. This source uses a new magnetic filter configuration. This magnetic filter is efficient in modifying the plasma potential distribution which can reduce the axial energy spread of the extracted ion beam. Energy spreads as low as 0.6 eV have been obtained. The electron temperature in this source has also been found to be about 0.1 eV. Furthermore, the new source configuration is capable of adjusting the radial plasma potential distribution which can improve the transverse ion energy, which results in a low beam emittance. The co-axial source can be used for a number of different applications such as ion projection lithography and radioactive ion beam projects.

  16. Co-axial multicusp source for low axial energy spread ion beam production

    Science.gov (United States)

    Lee, Y.; Gough, R. A.; Leung, K. N.; Vujic, J.; Williams, M. D.; Zahir, N.

    1999-09-01

    A co-axial multicusp ion source has been designed and tested. This source uses a new magnetic filter configuration. This magnetic filter is efficient in modifying the plasma potential distribution which can reduce the axial energy spread of the extracted ion beam. Energy spreads as low as 0.6 eV have been obtained. The electron temperature in this source has also been found to be about 0.1 eV. Furthermore, the new source configuration is capable of adjusting the radial plasma potential distribution which can improve the transverse ion energy, which results in a low beam emittance. The co-axial source can be used for a number of different applications such as ion projection lithography and radioactive ion beam projets.

  17. Axial flow heat exchanger devices and methods for heat transfer using axial flow devices

    Energy Technology Data Exchange (ETDEWEB)

    Koplow, Jeffrey P.

    2016-02-16

    Systems and methods described herein are directed to rotary heat exchangers configured to transfer heat to a heat transfer medium flowing in substantially axial direction within the heat exchangers. Exemplary heat exchangers include a heat conducting structure which is configured to be in thermal contact with a thermal load or a thermal sink, and a heat transfer structure rotatably coupled to the heat conducting structure to form a gap region between the heat conducting structure and the heat transfer structure, the heat transfer structure being configured to rotate during operation of the device. In example devices heat may be transferred across the gap region from a heated axial flow of the heat transfer medium to a cool stationary heat conducting structure, or from a heated stationary conducting structure to a cool axial flow of the heat transfer medium.

  18. 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.

  19. Axial Tomography from Digitized Real Time Radiography

    Science.gov (United States)

    Zolnay, A. S.; McDonald, W. M.; Doupont, P. A.; McKinney, R. L.; Lee, M. M.

    1985-01-18

    Axial tomography from digitized real time radiographs provides a useful tool for industrial radiography and tomography. The components of this system are: x-ray source, image intensifier, video camera, video line extractor and digitizer, data storage and reconstruction computers. With this system it is possible to view a two dimensional x-ray image in real time at each angle of rotation and select the tomography plane of interest by choosing which video line to digitize. The digitization of a video line requires less than a second making data acquisition relatively short. Further improvements on this system are planned and initial results are reported.

  20. Ankylosing Spondylitis versus Nonradiographic Axial Spondyloarthritis

    DEFF Research Database (Denmark)

    Glintborg, Bente; Sørensen, Inge J; Østergaard, Mikkel

    2017-01-01

    OBJECTIVE: To compare baseline disease activity and treatment effectiveness in biologic-naive patients with nonradiographic axial spondyloarthritis (nr-axSpA) and ankylosing spondylitis (AS) who initiate tumor necrosis factor inhibitor (TNFi) treatment and to study the role of potential confounders....../disease duration/TNFi-type/smoking/baseline disease activity) on TNFi adherence and response [e.g., Bath Ankylosing Spondylitis Activity Index (BASDAI) 50%/20 mm]. RESULTS: The study included 1250 TNFi-naive patients with axSpA (29% nr-axSpA, 50% AS, 21% lacked radiographs of sacroiliac joints). Patients...

  1. Tunable axial potentials for atom chip waveguides

    CERN Document Server

    Stickney, James A; Imhof, Eric; Kroese, Bethany R; Crow, Jonathon A R; Olson, Spencer E; Squires, Matthew B

    2014-01-01

    We present a method for generating algebraically precise magnetic potentials along the axis of a cold atom waveguide near the surface of an atom chip. With a single chip design consisting of several wire pairs, various axial potentials can be created, including double wells, triple wells, and pure harmonic traps with suppression of higher order terms. We characterize the error along a harmonic trap between the expected algebraic form and magnetic field simulations and find excel- lent agreement, particularly at small displacements from the trap center. Finally, we demonstrate experimental control over the bottom fields of an asymmetric double well potential.

  2. Composite Axial Flow Propulsor for Small Aircraft

    OpenAIRE

    2005-01-01

    This work focuses on the design of an axial flow ducted fan driven by a reciprocating engine. The solution minimizes the turbulization of the flow around the aircraft. The fan has a rotor - stator configuration. Due to the need for low weight of the fan, a carbon/epoxy composite material was chosen for the blades and the driving shaft.The fan is designed for optimal isentropic efficiency and free vortex flow. A stress analysis of the rotor blade was performed using the Finite Element  Method....

  3. Cervical Spine Axial Rotation Goniometer Design

    Directory of Open Access Journals (Sweden)

    Emin Ulaş Erdem

    2012-06-01

    Full Text Available To evaluate the cervical spine rotation movement is quiet harder than other joints. Configuration and arrangement of current goniometers and devices is not always practic in clinics and some methods are quiet expensive. The cervical axial rotation goniometer designed by the authors is consists of five pieces (head apparatus, chair, goniometric platform, eye pads and camera. With this goniometer design a detailed evaluation of cervical spine range of motion can be obtained. Besides, measurement of "joint position sense" which is recently has rising interest in researches can be made practically with this goniometer.

  4. Through-Flow Calculations in Axial Turbomachinery

    Science.gov (United States)

    1976-10-01

    downstzega of the effective throat which is displaced upstream away from its kominal plano flow •_stion. Test data .-n nigh deflection blading tested in...AXIAL PIE ANGs-L;- VrELOC.I- T/Y SI~NG,’ c (,o nd) 2 .959 00 SO~ IS~N1RO iCoWLE tNJ\\\\ NJ\\v ON45l~5INi +U~SWC E.AAINZN5W~N3~6 8Lk5~ P-tO RM~ C -5A

  5. 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

  6. Axial interbody arthrodesis of the L5-S1 segment: a systematic review of the literature.

    Science.gov (United States)

    Schroeder, Gregory D; Kepler, Christopher K; Vaccaro, Alexander R

    2015-09-01

    The object of this study was to determine the fusion rate and safety profile of an axial interbody arthrodesis of the L5-S1 motion segment. A systematic search of MEDLINE was conducted for literature published between January 1, 2000, and August 17, 2014. All peer-reviewed articles related to the fusion rate of L5-S1 and the safety profile of an axial interbody arthrodesis were evaluated. Seventy-four articles were identified, but only 15 (13 case series and 2 retrospective cohort studies) met the study inclusion criteria. The overall pseudarthrosis rate at L5-S1 was 6.9%, and the rate of all other complications was 12.9%. A total of 14.4% of patients required additional surgery, and the infection rate was 5.4%. Deformity studies reported a significantly increased rate of complications (46.3%), and prospectively collected data demonstrated significantly higher complication (36.8%) and revision (22.6%) rates. Lastly, studies with a conflict of interest reported lower complication rates (12.4%). A systematic review of the literature indicates that an axial interbody fusion performed at the lumbosacral junction is associated with a high fusion rate (93.15%) and an acceptable complication rate (12.90%). However, these results are based mainly on retrospective case series by authors with a conflict of interest. The limited prospective data available indicate that the actual fusion rate may be lower and the complication rate may be higher than currently reported.

  7. Axial residual stresses in boron fibers

    Science.gov (United States)

    Behrendt, D. R.

    1978-01-01

    A method of measuring axial residual stresses in boron fibers is presented. With this method, the axial residual stress distribution as a function of radius is determined from the fiber surface to the core including the average residual stress in the core. Such measurements on boron on tungsten (B/W) fibers show that the residual stresses for 102, 142, 203, and 366 micron diam fibers are similar, being compressive at the surface and changing monotonically to a region of tensile stress within the boron. At approximately 25% of the original radius, the stress reaches a maximum tensile stress of about 860 MN sq m and then decreases to a compressive stress near the tungsten boride core. Data are presented for 203-micron diam B/W fibers that show annealing above 900 C reduces the residual stresses. A comparison between 102-micron diam B/W and boron on carbon (B/C) show that the residual stresses are similar in the outer regions of the fibers, but that large differences near and in the core are observed. Fracture of boron fibers is discussed.

  8. Single Rod Vibration in Axial Flow

    Science.gov (United States)

    Weichselbaum, Noah; Wang, Shengfu; Bardet, Philippe

    2013-11-01

    Fluid structure interaction of a single rod in axial flow is a coupled dynamical system present in many application including nuclear reactors, steam generators, and towed antenna arrays. Fluid-structure response can be quantified thanks to detailed experimental data where both structure and fluid responses are recorded. Such datum deepen understanding of the physics inherent to the system and provide high-dimensionality quantitative measurements to validate coupled structural and CFD codes with various level of complexity. In this work, single rods fixed on both ends in a concentric pipe, are subjected to an axial flow with Reynolds number based on hydraulic diameter of Re =4000. Rods of varying material stiffness and diameter are utilized in the experiment resulting in a range of dimensionless U between 0.5 and 1, where U = (ρA/EI)1/2uL. Experimental measurements of the velocity field around the rod are taken with PIV from time-resolved Nd:YLF laser and a high speed CMOS camera. Three-dimensional and temporal vibration and deflection of the rod is recorded with shadowgraphy utilizing two sets of pulsed high power LED and dedicated CMOS camera. Through integration of these two diagnostics, it is possible to reconstruct the full FSI domain providing unique validation data.

  9. Multi-frequency axial transmission bone ultrasonometer.

    Science.gov (United States)

    Tatarinov, Alexey; Egorov, Vladimir; Sarvazyan, Noune; Sarvazyan, Armen

    2014-07-01

    The last decade has seen a surge in the development of axial transmission QUS (Quantitative UltraSound) technologies for the assessment of long bones using various modes of acoustic waves. The condition of cortical bones and the development of osteoporosis are determined by numerous mechanical, micro-structural, and geometrical or macro-structural bone properties like hardness, porosity and cortical thickness. Such complex manifestations of osteoporosis require the evaluation of multiple parameters with different sensitivities to the various properties of bone that are affected by the disease. This objective may be achieved by using a multi-frequency ultrasonic examination The ratio of the acoustic wavelength to the cortical thickness can be changed by varying the frequency of the ultrasonic pulse propagating through the long bone that results in the change in composition of the induced wave comprised of a set of numerous modes of guided, longitudinal, and surface acoustic waves. The multi-frequency axial transmission QUS method developed at Artann Laboratories (Trenton, NJ) is implemented in the Bone Ultrasonic Scanner (BUSS). In the current version of the BUSS, a train of ultrasonic pulses with 60, 100, 400, 800, and 1200 kHz frequencies is used. The developed technology was tested on a variety of bone phantoms simulating normal, osteopenic, and osteoporotic bones. The results of this study confirm the feasibility of the multi-frequency approach for the assessment of the processes leading to osteoporosis.

  10. AERODYNAMIC AND BLADING DESIGN OF MULTISTAGE AXIAL FLOW COMPRESSORS

    Science.gov (United States)

    Crouse, J. E.

    1994-01-01

    with fourth-degree polynomial functions of path distance from the maximum thickness point. Input to the aerodynamic and blading design program includes the annulus profile, the overall compressor mass flow, the pressure ratio, and the rotative speed. A number of input parameters are also used to specify and control the blade row aerodynamics and geometry. The output from the aerodynamic solution has an overall blade row and compressor performance summary followed by blade element parameters for the individual blade rows. If desired, the blade coordinates in the streamwise direction for internal flow analysis codes and the coordinates on plane sections through blades for fabrication drawings may be stored and printed. The aerodynamic and blading design program for multistage axial-flow compressors is written in FORTRAN IV for batch execution and has been implemented on an IBM 360 series computer with a central memory requirement of approximately 470K of 8 bit bytes. This program was developed in 1981.

  11. Study on the Axial Dispersion of Liquid in Column Flotation

    Institute of Scientific and Technical Information of China (English)

    周鵾; 曾爱武; 高长宝; 余国琮

    2003-01-01

    An experimental study on the axial dispersion of liquid was carried out in a 0.382-m-ID flotation column packed with different structured packings or free of packings. The correlations of axial Peclet numbers with the liquid and gas superficial Reynolds numbers were developed for various packings. Among the packings tested, it is found that in the column packed with 250Y or 350Y packings the axial dispersion is the lowest. The addition of frother can decrease the axial dispersion. By the simulation analysis of the one-dimension dispersion model of packed flotation column, it is found that small axial dispersion, high collection rate constant and low axial liquid velocity can increase the collection zone recovery.

  12. Analysis of axial fan noise with the help of the Lowson formalism

    Science.gov (United States)

    Bridelance, J. P.

    1980-11-01

    Slow rotation velocity, high lift profile blades are studied as a means of reducing specific acoustic power, shown in empirical studies of axial fans and helicopter blades to be highly correlated with rotational speed. Results for a new blade design are compared with classic axial fans for flow, pressure, efficiency, and noise. The Lowson method is used to extend Lighthill's (1952) work on acoustic fields generated by constant speed rotors, to rotors under acceleration. The source is considered as a point on the blades undergoing periodic fluctuations due to upstream turbulence, definable by a Fourier series, and the acoustic emissions result from unstable changes whose amplitude depends on the order of the Bessel function.

  13. Effect of Radial Density Configuration on Wave Field and Energy Flow in Axially Uniform Helicon Plasma

    Science.gov (United States)

    Chang, Lei; Li, Qingchong; Zhang, Huijie; Li, Yinghong; Wu, Yun; Zhang, Bailing; Zhuang, Zhong

    2016-08-01

    The effect of the radial density configuration in terms of width, edge gradient and volume gradient on the wave field and energy flow in an axially uniform helicon plasma is studied in detail. A three-parameter function is employed to describe the density, covering uniform, parabolic, linear and Gaussian profiles. It finds that the fraction of power deposition near the plasma edge increases with density width and edge gradient, and decays in exponential and “bump-on-tail” profiles, respectively, away from the surface. The existence of a positive second-order derivative in the volume density configuration promotes the power deposition near the plasma core, which to our best knowledge has not been pointed out before. The transverse structures of wave field and current density remain almost the same during the variation of density width and gradient, confirming the robustness of the m=1 mode observed previously. However, the structure of the electric wave field changes significantly from a uniform density configuration, for which the coupling between the Trivelpiece-Gould (TG) mode and the helicon mode is very strong, to non-uniform ones. The energy flow in the cross section of helicon plasma is presented for the first time, and behaves sensitive to the density width and edge gradient but insensitive to the volume gradient. Interestingly, the radial distribution of power deposition resembles the radial profile of the axial component of current density, suggesting the control of the power deposition profile in the experiment by particularly designing the antenna geometry to excite a required axial current distribution. supported by National Natural Science Foundation of China (No. 11405271)

  14. High-Frequency Axial Fatigue Test Procedures for Spectrum Loading

    Science.gov (United States)

    2016-07-20

    REPORT NO: NAWCADPAX/TIM-2016/49 HIGH - FREQUENCY AXIAL FATIGUE TEST PROCEEDURES FOR SPECTRUM LOADING by David T. Rusk, AIR...OF THE NAVY NAVAL AIR WARFARE CENTER AIRCRAFT DIVISION PATUXENT RIVER, MARYLAND NAWCADPAX/TIM-2016/49 20 July 2016 HIGH - FREQUENCY AXIAL...Technical Information Memorandum 3. DATES COVERED 4. TITLE AND SUBTITLE High - Frequency Axial Fatigue Test Procedures for Spectrum Loading

  15. A technique to determine a desired preparation axial inclination.

    Science.gov (United States)

    Parker, M Harry; Ivanhoe, John R; Blalock, John S; Frazier, Kevin B; Plummer, Kevin D

    2003-10-01

    The guidelines recommended in the literature for the convergence angle of a crown preparation vary from 3 to 24 degrees. There is a lack of guidelines on techniques to achieve a specific axial inclination. The purpose of this article was to present a practical technique, with a diamond rotary cutting instrument of known axial inclination, to determine the diamond rotary cutting instrument angulations required to achieve the desired axial inclination of a preparation.

  16. Design and Test of a Transonic Axial Splittered Rotor

    Science.gov (United States)

    2015-06-15

    AXIAL SPLITTERED ROTOR A new design procedure was developed that uses commercial-off-the-shelf software (MATLAB, SolidWorks , and ANSYS-CFX) for the...TRANSONIC AXIAL SPLITTERED ROTOR Report Title A new design procedure was developed that uses commercial-off-the-shelf software (MATLAB, SolidWorks , and...that uses commercial-off-the-shelf software (MATLAB, SolidWorks , and ANSYS-CFX) for the geometric rendering and analysis of a transonic axial

  17. Rotordynamics of Turbine Labyrinth Seals with Rotor Axial Shifting

    OpenAIRE

    Jinxiang Xi; Rhode, David L.

    2006-01-01

    Rotors in high-performance steam turbines experience a significant axial shifting during starting and stopping processes due to thermal expansion, for example. This axial shifting could significantly alter the flow pattern and the flow-induced rotordynamic forces in labyrinth seals, which in turn, can considerably affect the rotor-seal system performance. This paper investigates the influence of the rotor axial shifting on leakage rate as well as rotordynamic forces in hi...

  18. Velocity profiles in strongly turbulent Taylor-Couette flow

    CERN Document Server

    Grossmann, Siegfried; Sun, Chao

    2013-01-01

    We derive the velocity profiles in strongly turbulent Taylor-Couette flow for the general case of independently rotating cylinders. The theory is based on the Navier-Stokes equations in the appropriate (cylinder) geometry. In particular, we derive the axial and the angular velocity profiles as functions of distance from the cylinder walls and find that both follow a logarithmic profile, with downwards-bending curvature corrections, which are more pronounced for the angular velocity profile as compared to the axial velocity profile, and which strongly increase with decreasing ratio $\\eta$ between inner and outer cylinder radius. In contrast, the azimuthal velocity does not follow a log-law. We then compare the angular and azimuthal velocity profiles with the recently measured profiles in the ultimate state of (very) large Taylor numbers. Though the {\\em qualitative} trends are the same -- down-bending for large wall distances and (properly shifted and non-dimensionalized) angular velocity profile $\\omega^+(r)$...

  19. Modular functional organisation of the axial locomotor system in salamanders.

    Science.gov (United States)

    Cabelguen, Jean-Marie; Charrier, Vanessa; Mathou, Alexia

    2014-02-01

    Most investigations on tetrapod locomotion have been concerned with limb movements. However, there is compelling evidence that the axial musculoskeletal system contributes to important functions during locomotion. Adult salamanders offer a remarkable opportunity to examine these functions because these amphibians use axial undulations to propel themselves in both aquatic and terrestrial environments. In this article, we review the currently available biological data on axial functions during various locomotor modes in salamanders. We also present data showing the modular organisation of the neural networks that generate axial synergies during locomotion. The functional implication of this modular organisation is discussed.

  20. Energy Dissipation in Sandwich Structures During Axial Compression

    DEFF Research Database (Denmark)

    Urban, Jesper

    2002-01-01

    The purpose of this paper is to investigate the energy dissipation in sandwich structures during axial crushing. Axial crushing tests on six sandwich elements are described. The sandwich elements consist of a polyurethane core and E-glass/Polyester skin. The elements compare to full-scale structu......The purpose of this paper is to investigate the energy dissipation in sandwich structures during axial crushing. Axial crushing tests on six sandwich elements are described. The sandwich elements consist of a polyurethane core and E-glass/Polyester skin. The elements compare to full...

  1. Numerical optimisation of an axial turbine; Numerische Optimierung einer Axialturbine

    Energy Technology Data Exchange (ETDEWEB)

    Welzel, B.

    1998-12-31

    The author presents a method for automatic shape optimisation of components with internal or external flow. The method combines a program for numerical calculation of frictional turbulent flow with an optimisation algorithm. Algorithms are a simplex search strategy and an evolution strategy. The shape of the component to be optimized is variable due to shape parameters modified by the algorithm. For each shape, a flow calculation is carried out on whose basis a functional value like performance, loss, lift or resistivity is calculated. For validation, the optimisation method is used in simple examples with known solutions. It is applied. It is applied to the components of a slow-running axial turbine. Components with accelerated and delayed rotationally symmetric flow and 2D blade profiles are optimized. [Deutsch] Es wird eine Methode zur automatischen Formoptimierung durchstroemter oder umstroemter Bauteile vorgestellt. Diese koppelt ein Programm zur numerischen Berechnung reibungsbehafteter turbulenter Stroemungen mit einem Optimierungsalgorithmus. Dabei kommen als Algorithmen eine Simplex-Suchstrategie und eine Evolutionsstrategie zum Einsatz. Die Form des zu optimierenden Koerpers ist durch Formparameter, die vom Algorithmus veraendert werden, variabel. Fuer jede Form wird eine Stroemungsberechnung durchgefuehrt und mit dieser ein Funktionswert wie Wirkungsgrad, Verlust, Auftrieb oder Widerstandskraft berechnet. Die Optimierungsmethode wird zur Validierung in einfachen Beispielen mit bekannter Loesung eingesetzt. Zur Anwendung kommt sie in den einzelnen Komponenten einer langsamlaeufigen Axialturbine. Es werden Bauteile mit beschleunigter und verzoegerter rotationssymmetrischer Stroemung und 2D-Schaufelprofile optimiert. (orig.)

  2. Aerodynamic Modelling and Optimization of Axial Fans

    DEFF Research Database (Denmark)

    Sørensen, Dan Nørtoft

    A numerically efficient mathematical model for the aerodynamics oflow speed axial fans of the arbitrary vortex flow type has been developed.The model is based on a blade-element principle, whereby therotor is divided into a number of annular streamtubes.For each of these streamtubes relations...... for velocity, pressure andradial position are derived from the conservationlaws for mass, tangential momentum and energy.The resulting system of equations is non-linear and, dueto mass conservation and pressure equilibrium far downstream of the rotor,strongly coupled.The equations are solved using the Newton...... distributionsof pitch angle and chord length have been chosen as independent variablesin the optimizations.Besides restricting the geometry of the rotor,constraints have been added to ensure a required pressure rise as well asnon-stalled flow conditions.Optimizations have been performed tomaximize the mean value...

  3. The Axial Part Phrase in Japanese

    Directory of Open Access Journals (Sweden)

    Kaori Takamine

    2007-01-01

    Full Text Available In this paper, I investigate the categorial status of spatial terms in locative/directional expressions in Japanese. I will show that a certain class of spatial terms have a distinct categorial status from both regular postpositions and nouns. On one hand, syntactic diagnostics such as doubling, coordination by to, and co-occurrence with demonstratives indicate that these spatial terms belong to a nominal category rather than to a postpositional category. On the other hand, the fact that these spatial terms are modified by range modifiers indicates that they are more similar to regular postpositions than to nouns. On the basis of these diagnostics, I will argue that spatial terms in Japanese need to be assigned a new category Axial Part Phrase which is proposed by Svenonius 2006.

  4. Composite Axial Flow Propulsor for Small Aircraft

    Directory of Open Access Journals (Sweden)

    R. Poul

    2005-01-01

    Full Text Available This work focuses on the design of an axial flow ducted fan driven by a reciprocating engine. The solution minimizes the turbulization of the flow around the aircraft. The fan has a rotor - stator configuration. Due to the need for low weight of the fan, a carbon/epoxy composite material was chosen for the blades and the driving shaft.The fan is designed for optimal isentropic efficiency and free vortex flow. A stress analysis of the rotor blade was performed using the Finite Element  Method. The skin of the blade is calculated as a laminate and the foam core as a solid. A static and dynamic analysis were made. The RTM technology is compared with other technologies and is described in detail. 

  5. Aerodynamics of advanced axial-flow turbomachinery

    Science.gov (United States)

    Serovy, G. K.; Kavanagh, P.; Kiishi, T. H.

    1980-01-01

    A multi-task research program on aerodynamic problems in advanced axial-flow turbomachine configurations was carried out at Iowa State University. The elements of this program were intended to contribute directly to the improvement of compressor, fan, and turbine design methods. Experimental efforts in intra-passage flow pattern measurements, unsteady blade row interaction, and control of secondary flow are included, along with computational work on inviscid-viscous interaction blade passage flow techniques. This final report summarizes the results of this program and indicates directions which might be taken in following up these results in future work. In a separate task a study was made of existing turbomachinery research programs and facilities in universities located in the United States. Some potentially significant research topics are discussed which might be successfully attacked in the university atmosphere.

  6. Casimir Energy in the Axial Gauge

    CERN Document Server

    Esposito, G; Kirsten, K; Esposito, Giampiero; Kamenshchik, Alexander Yu.; Kirsten, Klaus

    2000-01-01

    The zero-point energy of a conducting spherical shell is studied by imposing the axial gauge via path-integral methods, with boundary conditions on the electromagnetic potential and ghost fields. The coupled modes are then found to be the temporal and longitudinal modes for the Maxwell field. The resulting system can be decoupled by studying a fourth-order differential equation with boundary conditions on longitudinal modes and their second derivatives. The exact solution of such equation is found by using a Green-function method, and is obtained from Bessel functions and definite integrals involving Bessel functions. Complete agreement with a previous path-integral analysis in the Lorenz gauge, and with Boyer's value, is proved in detail.

  7. Aerodynamic modelling and optimization of axial fans

    Energy Technology Data Exchange (ETDEWEB)

    Noertoft Soerensen, Dan

    1998-01-01

    A numerically efficient mathematical model for the aerodynamics of low speed axial fans of the arbitrary vortex flow type has been developed. The model is based on a blade-element principle, whereby the rotor is divided into a number of annular stream tubes. For each of these stream tubes relations for velocity, pressure and radial position are derived from the conservation laws for mass, tangential momentum and energy. The equations are solved using the Newton-Raphson methods, and solutions converged to machine accuracy are found at small computing costs. The model has been validated against published measurements on various fan configurations, comprising two rotor-only fan stages, a counter-rotating fan unit and a stator-rotor stator stage. Comparisons of local and integrated properties show that the computed results agree well with the measurements. Optimizations have been performed to maximize the mean value of fan efficiency in a design interval of flow rates, thus designing a fan which operates well over a range of different flow conditions. The optimization scheme was used to investigate the dependence of maximum efficiency on 1: the number of blades, 2: the width of the design interval and 3: the hub radius. The degree of freedom in the choice of design variable and constraints, combined with the design interval concept, provides a valuable design-tool for axial fans. To further investigate the use of design optimization, a model for the vortex shedding noise from the trailing edge of the blades has been incorporated into the optimization scheme. The noise emission from the blades was minimized in a flow rate design point. Optimizations were performed to investigate the dependence of the noise on 1: the number of blades, 2: a constraint imposed on efficiency and 3: the hub radius. The investigations showed, that a significant reduction of noise could be achieved, at the expense of a small reduction in fan efficiency. (EG) 66 refs.

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V. [Inst. of Radiobiology, Minsk Univ. (Belarus); Boulyga, S.F. [Inst. of Inorganic Chemistry and Analytical Chemistry, Johannes Gutenberg-Univ. Mainz, Mainz (Germany); Becker, J.S. [Central Div. of Analytical Chemistry, Research Centre Juelich, Juelich (Germany)

    2005-07-01

    An analytical method is described for the estimation of uranium concentrations, of {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10{sup -9}g/g to 2.0 x 10{sup -6}g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and the average value amounted to 9.4{+-}0.3 MWd/(kg U). (orig.)

  10. 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.

  11. 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

  12. Analysis of Axial Flow Ventilation Fans by Vortex - Method.

    Science.gov (United States)

    Hardin, Richard Anthony

    A steady vortex-lattice method is used to solve the lifting surface equation for an axial flow fan. The type of fan studied is designed for industrial and ventilation applications and in thermofluid systems such as cooling towers. The fan blades are thin cambered surfaces manufactured from metal sheets. The numerical approach is inviscid and results in a boundary value problem with viscous effects partially accounted for by application of drag coefficient data. A non-linear wake alignment procedure is used to account for the effects of vorticity shedding in the wake and variation in wake geometry with operating conditions. The wake alignment procedure is semi-free with wake input parameters required for accurate use of the technique. A study of the wake parameters was conducted and gave trends in the variation of their values with flow rate. At "free-air" conditions, flow visualization estimates of these parameters were found to agree with those from the computations. Comparisons are made between the measured and predicted fan performance with and without a surrounding duct. The comparison of the results were especially good at the "free-air" condition using wake parameters determined from flow visualization and an inlet velocity profile measured using hot-wire anemometry. To enable better understanding of basic flow phenomena and to provide data for verification of numerical analyses, a method for measuring unsteady surface pressure on a rotating axial-flow fan blade was devised. Unsteadiness of pressure on the blade surfaces is due to the effects of upstream fan motor supports and other installation features. A pressure transducer and signal amplification circuit were mounted on a circuit board at the rotating hub with signals taken off the rotating shaft through copper disk-mercury slip rings. The pressure difference across the blade was determined and the data were corrected for time lag and distortion caused by the length of tubing. The pressure difference

  13. Through flow analysis within axial flow turbomachinery blade rows

    Science.gov (United States)

    Girigoswami, H.

    1986-09-01

    Using Katsanis' Through Flow Code, inviscid flow through an axial flow compressor rotor blade as well as flow through inlet guide vanes are analyzed and the computed parameters such as meridional velocity distribution, axial velocity distribution along radial lines, and velocity distribution over blade surfaces are presented.

  14. Improved axial position detection in optical tweezers measurements

    DEFF Research Database (Denmark)

    Dreyer, Jakob Kisbye; Berg-Sørensen, Kirstine; Oddershede, Lene

    2004-01-01

    We investigate the axial position detection of a trapped microsphere in an optical trap by using a quadrant photodiode. By replacing the photodiode with a CCD camera, we obtain detailed information on the light scattered by the microsphere. The correlation of the interference pattern with the axial...

  15. Testing of Axially Loaded Bucket Foundation with Applied Overburden Pressure

    DEFF Research Database (Denmark)

    Vaitkunaite, Evelina; Ibsen, Lars Bo; Nielsen, Benjaminn Nordahl

    This report analyses laboratory testing data performed with a bucket foundation model subjected to axial loading. The examinations were conducted at the Geotechnical laboratory of Aalborg University. The report aims at showing and discussing the results of the static and cyclic axial loading tests...

  16. Test Setup for Axially Loaded Piles in Sand

    DEFF Research Database (Denmark)

    Thomassen, Kristina

    The test setup for testing axially static and cyclic loaded piles in sand is described in the following. The purpose for the tests is to examine the tensile capacity of axially loaded piles in dense fully saturated sand. The pile dimensions are chosen to resemble full scale dimension of piles used...... in offshore pile foundations today....

  17. An Unbroken Axial-Vector Current Conservation Law

    Science.gov (United States)

    Sharafiddinov, Rasulkhozha S.

    2016-04-01

    The mass, energy and momentum of the neutrino of a true flavor have an axial-vector nature. As a consequence, the left-handed truly neutral neutrino in an axial-vector field of emission can be converted into a right-handed one and vice versa. This predicts the unidenticality of masses, energies and momenta of neutrinos of the different components. Recognizing such a difference in masses, energies, momenta and accepting that the left-handed axial-vector neutrino and the right-handed antineutrino of true neutrality refer to long-lived C-odd leptons, and the right-handed truly neutral neutrino and the left-handed axial-vector antineutrino are of short-lived fermions of C-oddity, we would write a new CP-even Dirac equation taking into account the flavor symmetrical axial-vector mass, energy and momentum matrices. Their presence explains the spontaneous mirror symmetry violation, confirming that an axial-vector current conservation law has never violated. They reflect the availability of a mirror Minkowski space in which a neutrino is characterized by left as well as by right space-time coordinates. Therefore, it is not surprising that whatever the main purposes experiments about a quasielastic axial-vector mass say in favor of an axial-vector mirror Minkowski space-time.

  18. A new approach to radial and axial gauges

    Science.gov (United States)

    Weigert, Heribert; Heinz, Ulrich

    1992-03-01

    We develop a new path integral formulation of QCD in radial and axial gauges. This formalism yields free propagators which are free of gauge poles. We find that radial gauges are ghost free. In axial gauges ghosts cannot generally be excluded from the formalism due to the need to fix the residual gauge freedom.

  19. Design and Analysis of an Axially Laminated Reluctance Motor for Variable-Speed Applications

    Directory of Open Access Journals (Sweden)

    BESER, E. K.

    2013-02-01

    Full Text Available In this paper, an axially laminated reluctance motor is presented. First, a set of a finite element analysis (FEA on three different axially laminated rotor geometries was carried out and torque profiles of the rotors were predicted. The effect of the stator slot skewing on the torque profiles were also examined in the analysis. After deciding the rotor geometry, the mathematical model of the proposed motor was formed in terms of a,b,c variables and simulations were performed. Motor prototype and motor drive were introduced. Torque profiles of the motor were measured for different current values and load test were realized. Experimental results were compared to analysis and simulation results. There is a good accordance between experimental and simulation results. When the proposed motor is operated with electrical 120? mode as a brushless DC motor, the torque versus speed characteristic shows a DC series motor characteristic and speed of the motor can be easily controlled by regulating the bus voltage. These features make the proposed motor convenient for variable-speed applications such as electrical vehicles.

  20. Velocity profiles in strongly turbulent Taylor-Couette flow

    NARCIS (Netherlands)

    Grossmann, S.; Lohse, D.; Sun, C.

    2014-01-01

    We derive the velocity profiles in strongly turbulent Taylor-Couette flow for the general case of independently rotating cylinders. The theory is based on the Navier-Stokes equations in the appropriate (cylinder) geometry. In particular, we derive the axial and the angular velocity profiles as funct

  1. Axial vibration analysis of nanocones based on nonlocal elasticity theory

    Institute of Scientific and Technical Information of China (English)

    Shu-Qi Guo; Shao-Pu Yang

    2012-01-01

    Carbon nanocones have quite fascinating electronic and structural properties,whose axial vibration is seldom investigated in previous studies.In this paper,based on a nonlocal elasticity theory,a nonuniform rod model is applied to investigate the small-scale effect and the nonuniform effect on axial vibration of nanocones.Using the modified Wentzel-Brillouin-Kramers (WBK) method,an asymptotic solution is obtained for the axial vibration of general nonuniform nanorods.Then,using similar procedure,the axial vibration of nanocones is analyzed for nonuniform parameters,mode number and nonlocal parameters.Explicit expressions are derived for mode frequencies of clamped-clamped and clamped-free boundary conditions.It is found that axial vibration frequencies are highly overestimated by the classical rod model because of ignorance of the effect of small length scale.

  2. On Stationary Axially Symmetric Solutions in Brans-Dicke Theory

    CERN Document Server

    Kirezli, Pınar

    2015-01-01

    Stationary axially symmetric Brans-Dicke-Maxwell solutions are re-examined in the framework of the Brans-Dicke theory. We see that, employing a particular parametrization of the standard axially symmetric metric simplifies the procedure of obtaining the Ernst equations for axially symmetric electro-vacuum space-times for this theory. This analysis also permit us to construct a two parameter extension in both Jordan and Einstein frames of an old solution generating technique frequently used to construct axially symmetric solutions for Brans-Dicke theory from a seed solution of General Relativity. As applications of this technique, several known and new solutions are constructed including a general axially symmetric BD-Maxwell solution of Plebanski-Demianski with vanishing cosmological constant, i.e. the Kinnersley solution and general magnetized Kerr-Newman type solutions. Some physical properties and circular motion of test particles for a particular subclass of Kinnersley solution, i.e. Kerr-Newman-NUT type ...

  3. Flow control in axial fan inlet guide vanes by synthetic jets

    OpenAIRE

    Wurst P.; Trávníček Z.; Cyrus V.; Kordík J.

    2013-01-01

    Tested high pressure axial flow fan with hub/tip ratio of 0.70 and external diameter of 600 mm consisted of inlet guide vanes (IGV), rotor and stator blade rows. Fan peripheral velocity was 47 m/s. Air volume flow rate was changed by turning of rear part of the inlet guide vanes. At turning of 20 deg the flow was separated on the IGV profiles. The synthetic jets were introduced through radial holes in machine casing in the location before flow separation origin. Synthetic jet actuator was des...

  4. Stability Analysis of Restricted Axial Symmetric Collapse in $f(R)$ Gravity

    CERN Document Server

    Sharif, M

    2014-01-01

    In this paper, we analyze stability regions of a non-static restricted class of axially symmetric spacetime with anisotropic matter distribution. We develop dynamical as well as collapse equations within the framework of proposed perturbation technique for $f(R)=R+{\\epsilon}R^2$ model and explore dynamical instabilities at Newtonian and post-Newtonian regimes. It is concluded that radial profile of physical parameters like pressure anisotropy, energy density and higher curvature terms of the $f(R)$ model affect the instability ranges.

  5. Analysis of the axial electric field in a plasma-loaded-helix travelling wave tube

    Institute of Scientific and Technical Information of China (English)

    Xie Hong-Quan; Liu Pu-Kun

    2006-01-01

    A helix type slow wave structure filled with plasma is immersed in a strong longitudinal magnetic field. Taking into account the effect of the plasma and the dielectric, the system is separated radially into three regions. By means of the sheath model and Maxwell equation, the distribution of the electromagnetic field is established. Using the boundary conditions of each region, the dispersion relation of the slow wave structure is derived. The trend of change for the radial profile of the axial electric field is analysed respectively in different plasma densities, plasma column radius and dielectric constant by numerical computation. Some useful results are obtained on the basis of the discussion.

  6. Dynamic visualization of axial p-n junctions in single gallium nitride nanorods under electrical bias.

    Science.gov (United States)

    Lu, Yu-Jung; Lu, Ming-Yen; Yang, Yu-Chen; Chen, Hung-Ying; Chen, Lih-Juann; Gwo, Shangjr

    2013-09-24

    We demonstrate a direct visualization method based on secondary electron (SE) imaging in scanning electron microscopy for mapping electrostatic potentials across axial semiconductor nanorod p-n junctions. It is found that the SE doping contrast can be directly related to the spatial distribution of electrostatic potential across the axial nanorod p-n junction. In contrast to the conventional SE doping contrast achieved for planar p-n junctions, the quasi-one-dimensional geometry of nanorods allows for high-resolution, versatile SE imaging under high accelerating voltage, long working distance conditions. Furthermore, we are able to delineate the electric field profiles across the axial nanorod p-n junction as well as depletion widths at different reverse biases. By using standard p-n junction theory and secondary ion mass spectroscopy, the carrier concentrations of p- and n-regions can be further extracted from the depletion widths under reverse biasing conditions. This direct imaging method enables determination of electrostatic potential variation of p-n junctions in semiconductor nanorod and nanowire devices with a spatial resolution better than 10 nm.

  7. Axial Vibration Confinement in Nonhomogenous Rods

    Directory of Open Access Journals (Sweden)

    S. Choura

    2005-01-01

    Full Text Available A design methodology for the vibration confinement of axial vibrations in nonhomogenous rods is proposed. This is achieved by a proper selection of a set of spatially dependent functions characterizing the rod material and geometric properties. Conditions for selecting such properties are established by constructing positive Lyapunov functions whose derivative with respect to the space variable is negative. It is shown that varying the shape of the rod alone is sufficient to confine the vibratory motion. In such a case, the vibration confinement requires that the eigenfunctions be exponentially decaying functions of space, where the notion of spatial domain stability is introduced as a concept dual to that of the time domain stability. It is also shown that vibration confinement can be produced if the rod density and/or stiffness are varied with respect to the space variable while the cross-section area is kept constant. Several case studies, supporting the developed conditions imposed on the spatially dependent functions for vibration confinement in vibrating rods, are discussed. Because variation in the geometric and material properties might decrease the critical buckling loads, we also discuss the buckling problem.

  8. Computerized axial tomography in clinical pediatrics.

    Science.gov (United States)

    McCullough, D C; Kufta, C; Axelbaum, S P; Schellinger, D

    1977-02-01

    Computerized axial tomography (CAT), a noninvasive radiologie method, provides a new dimension in screening and diagnosis of intracranial pathology. Evaluation of 725 scans in infants and children demonstrates that CAT may be performed with negligible risk, although sedation and restraint are essential to the successful performance of studies in children under 6 years of age. CAT is the preferred initial diagnostic method in suspected hydrocephalls and is accurate in the detection and precise localization of brain tumors. The management of hydrocephalus and brain tumors has been significantly altered by the availability of CAT. Few invasive neuroradiologic procedures are required and pneumography is especially curtailed. Serial scanning is the best available method of monitoring ventricular alterations in hydrocephalus, tumor size during radiotherapy or chemotherapy, and postoperative recurrence of benign neoplasms. Complex intracranial anomalies are detectable with computerized tomography, but complete definition of pathology often requires angiography and air studies. Limited clinical experience in detecting neonatal intraventricular hemorrhage suggests that CAT will be a valuable tool for futlre investigations of that problem.

  9. Multimode interaction in axially excited cylindrical shells

    Directory of Open Access Journals (Sweden)

    Silva F. M. A.

    2014-01-01

    Full Text Available Cylindrical shells exhibit a dense frequency spectrum, especially near the lowest frequency range. In addition, due to the circumferential symmetry, frequencies occur in pairs. So, in the vicinity of the lowest natural frequencies, several equal or nearly equal frequencies may occur, leading to a complex dynamic behavior. So, the aim of the present work is to investigate the dynamic behavior and stability of cylindrical shells under axial forcing with multiple equal or nearly equal natural frequencies. The shell is modelled using the Donnell nonlinear shallow shell theory and the discretized equations of motion are obtained by applying the Galerkin method. For this, a modal solution that takes into account the modal interaction among the relevant modes and the influence of their companion modes (modes with rotational symmetry, which satisfies the boundary and continuity conditions of the shell, is derived. Special attention is given to the 1:1:1:1 internal resonance (four interacting modes. Solving numerically the governing equations of motion and using several tools of nonlinear dynamics, a detailed parametric analysis is conducted to clarify the influence of the internal resonances on the bifurcations, stability boundaries, nonlinear vibration modes and basins of attraction of the structure.

  10. Axially Symmetric Post-Newtonian Stellar Systems

    Directory of Open Access Journals (Sweden)

    Camilo Akímushkin

    2010-06-01

    Full Text Available We introduce a method to obtain self-consistent, axially symmetric disklike stellar models in the first post-Newtonian (1PN approximation. By using in the field equations of the 1PN approximation a distribution function (DF corresponding to a Newtonian model, two fundamental equations determining the 1PN corrections are obtained. The rotation curves of the corrected models differs from the classical ones and the corrections are clearly appreciable with values of the mass and radius of a typical galaxy. On the other hand, the relativistic mass correction can be ignored for all models. Resumen. Presentamos un método para obtener modelos estelares discoidales, axialmente simétricos, auto-consistentes en la primera aproximación post-Newtoniana (1PN. Usando en las ecuaciones de campo de la aproximación 1PN una función de distribución conocida (DF que corresponde a un modelo Newtoniano, se obtienen dos ecuaciones fundamentales para determinar las correcciones 1PN. Las curvas de rotación de los modelos corregidos difieren de las clásicas y las correcciones son claramente apreciables con los valores de la masa y el radio de una galaxia típica. Por otro lado, la corrección relativista de la masa se puede ignorar para todos los modelos.

  11. New treatment targets for axial spondyloarthritis.

    Science.gov (United States)

    Sieper, Joachim

    2016-12-01

    Axial spondyloarthritis (axSpA) patients can be divided into those with structural damage in the SI joint visible on X-rays, termed radiographic axSpA or AS, and those in an earlier phase of the disease, without structural damage in the SI joint, termed non-radiographic axSpA. TNF-blockers have been shown to be highly effective in the treatment of active axSpA. Interestingly, conventional DMARDs and also non-TNF-blocker biologics targeting IL-1, IL-6 and T cells (abatacept) are not effective. Recent interest has focused on the cytokines IL-23 and IL-17 as potential treatment targets in axSpA. An open-label trial with ustekinumab showed a good efficacy in AS patients. Two placebo-controlled phase 3 trials with a mAb blocking IL-17, secukinumab, showed a good reduction in disease activity, similar to that shown for TNF blockers. Probably triggered by inflammation, new bone formation is another hallmark in AS and a potentially important treatment target. However, a previously reported inhibitory effect of NSAID treatment could not be confirmed in a recent NSAID trial.

  12. Vertebrate Axial Patterning: From Egg to Asymmetry.

    Science.gov (United States)

    Houston, Douglas W

    2017-01-01

    The emergence of the bilateral embryonic body axis from a symmetrical egg has been a long-standing question in developmental biology. Historical and modern experiments point to an initial symmetry-breaking event leading to localized Wnt and Nodal growth factor signaling and subsequent induction and formation of a self-regulating dorsal "organizer." This organizer forms at the site of notochord cell internalization and expresses primarily Bone Morphogenetic Protein (BMP) growth factor antagonists that establish a spatiotemporal gradient of BMP signaling across the embryo, directing initial cell differentiation and morphogenesis. Although the basics of this model have been known for some time, many of the molecular and cellular details have only recently been elucidated and the extent that these events remain conserved throughout vertebrate evolution remains unclear. This chapter summarizes historical perspectives as well as recent molecular and genetic advances regarding: (1) the mechanisms that regulate symmetry-breaking in the vertebrate egg and early embryo, (2) the pathways that are activated by these events, in particular the Wnt pathway, and the role of these pathways in the formation and function of the organizer, and (3) how these pathways also mediate anteroposterior patterning and axial morphogenesis. Emphasis is placed on comparative aspects of the egg-to-embryo transition across vertebrates and their evolution. The future prospects for work regarding self-organization and gene regulatory networks in the context of early axis formation are also discussed.

  13. Nitinol stent design - understanding axial buckling.

    Science.gov (United States)

    McGrath, D J; O'Brien, B; Bruzzi, M; McHugh, P E

    2014-12-01

    Nitinol׳s superelastic properties permit self-expanding stents to be crimped without plastic deformation, but its nonlinear properties can contribute towards stent buckling. This study investigates the axial buckling of a prototype tracheobronchial nitinol stent design during crimping, with the objective of eliminating buckling from the design. To capture the stent buckling mechanism a computational model of a radial force test is simulated, where small geometric defects are introduced to remove symmetry and allow buckling to occur. With the buckling mechanism ascertained, a sensitivity study is carried out to examine the effect that the transitional plateau region of the nitinol loading curve has on stent stability. Results of this analysis are then used to redesign the stent and remove buckling. It is found that the transitional plateau region can have a significant effect on the stability of a stent during crimping, and by reducing the amount of transitional material within the stent hinges during loading the stability of a nitinol stent can be increased.

  14. Análisis de la fuerza axial en un transportador de sinfín // Analysis of axial force in a screw conveyor.

    Directory of Open Access Journals (Sweden)

    F. Aguilar Parés

    1999-01-01

    Full Text Available Durante el movimiento de un material en un transportador de sinfín surge una fuerza en dirección axial que influye en laselección de uno de los cojinetes de apoyo del equipo. En el artículo aparecen algunas soluciones constructivas que tienen encuentan la fuerza axial. Por otro lado se establece la relación entre la fuerza axial y el empuje axial y se precisa de quiendepende el sentido del empuje axial. Por último se propone un modelo matemático que relaciona la fuerza axial con la potenciarequerida por el equipo.Palabras claves: Transportador de sinf in, fuerza axial , empuje axial ._________________________________________________________________________AbstractDuring the movement of material in a screw conveyor surge a force in axial direction that influence in the selection of one ofthe equipment support bearings. Some constructive solutions appear in the article for considering the axial force. In the otherhand it is established the relation between axial force and axial thurst and it is precised whose direction thurst axial depend of.Finally it is proposed a mathematic model that relates the axial force with the power required by the equipment.Key words: Screw conveyor, axial force, axial thurst .

  15. Burn-up Function of Fuel Management Code for Aqueous Homogeneous Reactors and Its Validation%溶液堆物理计算程序FMCAHR燃耗功能及其验证

    Institute of Scientific and Technical Information of China (English)

    汪量子; 姚栋; 王侃

    2011-01-01

    介绍了FMCAHR程序的燃耗计算模型及流程,并使用燃耗基准题和DRAGON程序对燃耗计算结果进行验证.验证结果表明,FMCAHR燃耗计算功能的准确性较高,适用于溶液堆的燃耗计算分析.%Fuel Management Code for Aqueous Homogeneous Reactors(FMCAHR)is developed based on the Monte Carlo transport method,to analyze the physics characteristics of aqueous homogeneous reactors. FMCAHR has the ability of doing resonance treatment,searching for critical rod heights,thermal hydraulic parameters calculation,radiolytic-gas bubbles' calculation and burn-up calculation. This paper introduces the theory model and scheme of its bum-up function,and then compares its calculation results with benchmarks and with DRAGON'S burn-up results,which confirms its burn-up computing precision and its applicability in the burn-up calculation and analysis for aqueous solution reactors.

  16. Development of Burnup Calculation Function in Reactor Monte Carlo Code RMC%堆用蒙卡程序燃耗计算功能开发

    Institute of Scientific and Technical Information of China (English)

    佘顶; 王侃; 余纲林

    2012-01-01

    This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua university of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency.%堆用蒙卡程序(RMC)是由清华大学工程物理系REAL实验室自主开发的用于反应堆物理分析的中子输运蒙卡程序,本文主要介绍其燃耗计算功能的开发与验证.RMC的燃耗计算功能具有的特点:内部耦合ORIGEN,相比于外耦合方式,更加灵活和高效;使用基于能谱的单群截面统计方法,可在保证精度的前提下,显著提高计算效率;采取预估修正和中点近似等多种燃耗步策略,减小大燃耗步长时的计算误差.通过计算压水堆栅元、沸水堆组件、快堆等一系列基准题和算例,验证了RMC燃耗计算的正确性和速度优势.

  17. Active axial spondyloarthritis: potential role of certolizumab pegol

    Directory of Open Access Journals (Sweden)

    Ranatunga S

    2014-02-01

    Full Text Available Sriya Ranatunga, Anne V Miller Department of Internal Medicine, Southern Illinois University School of Medicine, Springfield, IL, USA Abstract: The axial spondyloarthropathies are a group of chronic inflammatory diseases that predominantly affect the axial joints. This group includes ankylosing spondylitis and nonradiographic axial spondyloarthropathy. While the pathogenesis of axial spondyloarthropathies is not clear, immunologically active tissues primarily include the entheses, ie, the areas where ligaments, tendons, and joint capsules attach to bone and to the annulus fibrosis at the vertebrae. One of the major mediators of the immune response in this group of diseases is tumor necrosis factor-alpha (TNFα. Blockade of TNFα results in reduced vascularity and inflammatory cell infiltration in the synovial tissues of affected joints. Certolizumab pegol (CZP is an Fc-free, PEGylated anti-TNFα monoclonal antibody. CZP has unique properties that differ from other available TNFα inhibitors by virtue of its lack of an Fc region, which minimizes potential Fc-mediated effects, and its PEGylation, which improves drug pharmacokinetics and bioavailability. It has been shown in clinical trials that CZP improves patient outcomes and reduces inflammation in the sacroiliac joints and spine in both ankylosing spondylitis and nonradiographic axial spondyloarthropathies. These data support CZP as a treatment option for axial spondyloarthropathies. Keywords: axial spondyloarthropathy, certolizumab pegol, anti-tumor necrosis factor-alpha, therapy

  18. Axial elongation following prolonged near work in myopes and emmetropes.

    Science.gov (United States)

    Woodman, Emily C; Read, Scott A; Collins, Michael J; Hegarty, Katherine J; Priddle, Scott B; Smith, Josephine M; Perro, Judd V

    2011-05-01

    To investigate the influence of a period of sustained near work upon axial length in groups of emmetropes (EMM) and myopes. Forty young adult subjects (20 myopes and 20 emmetropes) were recruited for the study. Myopes were further classified as early onset (EOM), late onset (LOM), stable (SM) or progressing (PM) subgroups. Axial length was measured with the IOLMaster instrument before, immediately after and then again 10 min after a continuous 30 min near task of 5 D accommodation demand. Measures of distance objective refraction were also collected. Significant changes in axial length were observed immediately following the near task. EOM axial length elongated on average by 0.027±0.021 mm, LOM by 0.014±0.020 mm, EMM by 0.010±0.015 mm, PM by 0.031±0.022 mm and SM by 0.014±0.018 mm. At the conclusion of the 10 min regression period, axial length measures were not significantly different from baseline values. Axial elongation was observed following a prolonged near task. Both EOM and PM groups showed increases in axial length that were significantly greater than emmetropes.

  19. Effectiveness of Adalimumab in Non-radiographic Axial Spondyloarthritis

    Science.gov (United States)

    Cantarini, Luca; Fabbroni, Marta; Talarico, Rosaria; Costa, Luisa; Caso, Francesco; Cuneo, Gian Luca; Frediani, Bruno; Faralli, Gabriele; Vitale, Antonio; Brizi, Maria Giuseppina; Sabadini, Luciano; Galeazzi, Mauro

    2015-01-01

    Abstract The primary aim of the study was to evaluate the long-term effectiveness of adalimumab (ADA) in a cohort of non-radiographic axial spondyloarthritis (nr-axSpA), and the secondary aims were to identify predictive factors of response and evaluate radiological progression. We evaluated 37 patients (male/female: 12/25; mean age 49 ± 14; mean disease duration: 6.3 ± 5.8) with active nr-axSpA (Assessment of SpondyloArthritis International Society criteria), despite the treatment with ≥1 nonsteroidal anti-inflammatory drug for at least 3 months, initiating the treatment with ADA 40 mg every other week. Patients were treated for 24 months, and evaluated at baseline, 6, 12, and 24 months. Outcome measures included Ankylosing Spondylitis Disease Activity Score, Bath Ankylosing Spondylitis Disease Activity Index (BASDAI), and Bath Ankylosing Spondylitis Functional Index. Radiograph of the spine and sacroiliac joints and magnetic resonance of the sacroiliac joints were performed at baseline and according to the standard of assessment for the disease. The proportion of patients that achieved a BASDAI50 response at 6, 12 and 24 months was 51.3%, 70.3%, and 76.8%, respectively. Treatment was well tolerated with no unexpected adverse events and/or serious adverse events. All patients remained on treatment for 2 years, with a good compliance. We did not identify any predictive factor of response to therapy. Moreover, modified Stoke Ankylosing Spondylitis Spine Score and Spondyloarthritis Research Consortium of Canada scores showed a trend of improvement during the study period. ADA was effective on clinical and radiological outcomes at 2-year follow-up; thus, early treatment with ADA may prevent radiographic damage and be associated with low disease activity or remission. Moreover, data from this cohort study have confirmed safety and tolerability profile of ADA in nr-axSpA in the long term. PMID:26222847

  20. Tensile Property of Bi-axial Warp Knitted Structure

    Institute of Scientific and Technical Information of China (English)

    沈为

    2003-01-01

    The tensile property of bi-axial warp knitted fabrics is tested and compared with that of the plain weave fabric. The results show that there are obvious differences between the tensile property of a bi-axial warp knitted fabric and that of a plain weave fabric.The former can give fuller play to the property of a high modulus yarn than the latter. The tensile strength of a bi-axial warp knitted fabric is linear with the number of yarns in the direction of force.

  1. The axial charges of the hidden-charm pentaquark states

    CERN Document Server

    Wang, Guang-Juan; Zhu, Shi-Lin

    2016-01-01

    With the chiral quark model, we have calculated the axial charges of the pentaquark states with $(I,I_3)=(\\frac{1}{2},\\frac{1}{2})$ and $J^{P}=\\frac{1}{2}^{\\pm},\\frac{3}{2}^{\\pm},\\frac{5}{2}^{\\pm}$. The $P_c$ states with the same $J^P$ quantum numbers but different color-spin-flavor configurations have very different axial charges, which encode important information on their underlying structures. For some of the $J^{P}=\\frac{3}{2}^{\\pm}$ or $\\frac{5}{2}^{\\pm}$ pentaquark states, their axial charges are much smaller than that of the proton.

  2. Dynamic Analysis of Axial Magnetic Forces for DVD Spindle Motors

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    The axial magnetic force, induced by the complicated flux linkage distribution from rotor magnet and stator slotted, is constructed by different relative heights and calculated by 3D finite element method (FEM) to analyze the dynamic characteristics for a DVD spindle motor. The axial magnetic force is designed to provide an axial stiffness and govern the natural frequency of the dynamic performance. According to the simulation results and experimental measurements, the dynamic behaviors are significantly improved with a variation of relative height of rotor magnet and stator slotted on a DVD spindle motor.

  3. Gravitational waves from the axial perturbations of hyperon stars

    Institute of Scientific and Technical Information of China (English)

    Wen De-Hua; Yan Jing; Liu Xue-Mei

    2012-01-01

    The eigen-frequencies of the axial w-mode oscillations of hyperon stars are examined.It is shown that as the appearance of hyperons softens the equation of state of the super-density matter,the frequency of gravitational waves from the axial w-mode of hyperon star becomes smaller than that of a traditional neutron star at the same stellar mass.Moreover,the eigenfrequencies of hyperon stars also have scaling universality.It is shown that the EURO thirdgeneration gravitational-wave detector has the potential to detect the gravitational-wave signal emitted from the axial w-mode oscillations of a hyperon star.

  4. Dispositivos de asistencia ventricular de tipo axial

    Directory of Open Access Journals (Sweden)

    Albert Miralles Cassina

    2009-04-01

    Full Text Available El uso de dispositivos de asistencia ventricular se ha ido extendiendo en las últimas décadas. La mejora de los resultados ha ido acompañada del diseño de nuevos aparatos más pequeños y eficientes como son las bombas de flujo axial. las características técnicas básicas de estos dispositivos se hallan en la generación de un flujo continuo unidireccional conseguido mediante sistemas de turbina que obtienen su alimentación de una fuente eléctrica. Las ventajas principales de estos dispositivos son: su facilidad de implantación por su tamaño reducido, su prolongada durabilidad gracias a su sencillo diseño y su eficiencia energética al utilizar energía eléctrica. Su utilidad se ha podido demostrar en diferentes indicaciones de asistencia circulatoria mecánica, como en el caso de puente al trasplante o dispositivo de recuperación miocárdica, si bien donde radica más interés es en su uso como sistemas de asistencia circulatoria mecánica definitiva. Existe una variedad de modelos de diversos diseños. Son dispositivos que permiten soporte normalmente univentricular izquierdo, que se implantan en el tórax con cánula de entrada a nivel ventricular y cánula de salida a nivel aórtico. En este artículo se efectúa una descripción de los principales sistemas disponibles en la actualidad, comentando las características técnicas, ventajas e inconvenientes y un resumen de la experiencia existente.

  5. Revealing atropisomer axial chirality in drug discovery.

    Science.gov (United States)

    LaPlante, Steven R; Edwards, Paul J; Fader, Lee D; Jakalian, Araz; Hucke, Oliver

    2011-03-07

    An often overlooked source of chirality is atropisomerism, which results from slow rotation along a bond axis due to steric hindrance and/or electronic factors. If undetected or not managed properly, this time-dependent chirality has the potential to lead to serious consequences, because atropisomers can be present as distinct enantiomers or diastereoisomers with their attendant different properties. Herein we introduce a strategy to reveal and classify compounds that have atropisomeric chirality. Energy barriers to axial rotation were calculated using quantum mechanics, from which predicted high barriers could be experimentally validated. A calculated rotational energy barrier of 20 kcal mol(-1) was established as a suitable threshold to distinguish between atropisomers and non-atropisomers with a prediction accuracy of 86%. This methodology was applied to subsets of drug databases in the course of which atropisomeric drugs were identified. In addition, some drugs were exposed that were not yet known to have this chiral attribute. The most valuable utility of this tool will be to predict atropisomerism along the drug discovery pathway. When used in concert with our compound classification scheme, decisions can be made during early discovery stages such as "hit-to-lead" and "lead optimization," to foresee and validate the presence of atropisomers and to exercise options of removing, further stabilizing, or rendering the chiral axis of interest more freely rotatable via SAR design, thereby decreasing this potential liability within a compound series. The strategy can also improve drug development plans, such as determining whether a drug or series should be developed as a racemic mixture or as an isolated single compound. Moreover, the work described herein can be extended to other chemical fields that require the assessment of potential chiral axes.

  6. The Design Method of Axial Flow Runners Focusing on Axial Flow Velocity Uniformization and Its Application to an Ultra-Small Axial Flow Hydraulic Turbine

    Directory of Open Access Journals (Sweden)

    Yasuyuki Nishi

    2016-01-01

    Full Text Available We proposed a portable and ultra-small axial flow hydraulic turbine that can generate electric power comparatively easily using the low head of open channels such as existing pipe conduits or small rivers. In addition, we proposed a simple design method for axial flow runners in combination with the conventional one-dimensional design method and the design method of axial flow velocity uniformization, with the support of three-dimensional flow analysis. Applying our design method to the runner of an ultra-small axial flow hydraulic turbine, the performance and internal flow of the designed runner were investigated using CFD analysis and experiment (performance test and PIV measurement. As a result, the runners designed with our design method were significantly improved in turbine efficiency compared to the original runner. Specifically, in the experiment, a new design of the runner achieved a turbine efficiency of 0.768. This reason was that the axial component of absolute velocity of the new design of the runner was relatively uniform at the runner outlet in comparison with that of the original runner, and as a result, the negative rotational flow was improved. Thus, the validity of our design method has been verified.

  7. The MOX fuel behaviour test IFA-597.4/.5/.6; Thermal and gas release data to a burn-up of 25 MWd/kgMOX

    Energy Technology Data Exchange (ETDEWEB)

    Tolonen, Pekka; Pihlatie, Mikko; Fujii, Hajime

    2001-02-15

    Responding to the needs of the member organisations, a research programme of MOX fuel has been established in the joint programme of the Halden Reactor Project. IFA-597.4, IFA- 597.5, and IFA-597.6, containing two MIMAS-MOX fuel rods, both equipped with a fuel centre thermocouple and a pressure bellows transducer, have been irradiated in the Halden Reactor since July 1997. The objective of the test series is study the thermal and fission gas release behaviour of MOX fuel, and to explore differences in performance between solid and hollow pellets. One of the rods has mainly solid pellets, while the other contains only hollow pellets. Both rods have an initial Pu-fissile enrichment of 6.07%. The cladding outside diameter is 9.5 mm, and the initial fuel-cladding gap 180 mum. The rods are being periodically uprated to study the fission gas release onset of MOX fuel. The first uprating was performed at approx 10 MWd/kgMOX resulting in significant gas release in both rods. In order to accumulate fission gas in the fuel matrix instead of releasing it, four UO{sub 2} rods were added to the rig at approx 13.5 MWd/kgMOX to suppress the linear heat rate of the MOX rods. Consequently, no gas release occurred during this low power operation. The UO{sub 2} rods were removed at approx 20.5 MWd/kgMOX, resulting in a power increase and significant gas release in both rods. The burnup of the rods has reached approx 25 MWd/kgMOX as of the end of October 2000. The following loading will operate at low power until late 2001 to avoid fission gas release. The target burnup of the test is 60 MWd/kgMOX. (Author)

  8. 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.

  9. The Nonradiographic Axial Spondyloarthritis, the Radiographic Axial Spondyloarthritis, and Ankylosing Spondylitis: The Tangled Skein of Rheumatology

    Science.gov (United States)

    Rawat, Roopa; Agrawal, Neha; Patil, Nilesh S.

    2017-01-01

    Since 1984 the diagnosis of ankylosing spondylitis (AS) has been based upon the modified New York (mNY) criteria with mandatory presence of radiographic sacroiliitis, without which the diagnosis is not tenable. However, it may take years or decades for radiographic sacroiliitis to develop delaying the diagnosis for long periods. It did not matter in the past because no effective treatment was available. However, with the availability of a highly effective treatment, namely, tumour necrosis factor-α inhibitors (TNFi), the issue of early diagnosis of AS acquired an urgency. The Assessment of SpondyloArthritis International Society (ASAS) classification criteria published in 2009 was a significant step towards this goal. These criteria described an early stage of the disease where sacroiliitis was demonstrable only on MRI but not on standard radiograph. Therefore, this stage of the disease was labelled “nonradiographic axial SpA” (nr-axSpA). But questions have been raised if, in search of early diagnosis, specificity was compromised. The Federal Drug Administration (FDA, USA) withheld approval for the use of TNFi in patients with nr-axSpA because of issues related to the specificity of these criteria. This review attempts to clarify some of these aspects of the nr-axSpA-AS relationship and also tries to answer the question whether ASAS classifiable radiographic axial spondyloarthritis (r-axSpA) term can be interchangeably used with the term AS. PMID:28555158

  10. Experimental investigation of the noise emission of axial fans under distorted inflow conditions

    Science.gov (United States)

    Zenger, Florian J.; Renz, Andreas; Becher, Marcus; Becker, Stefan

    2016-11-01

    An experimental investigation on the noise emission of axial fans under distorted inflow conditions was conducted. Three fans with forward-skewed fan blades and three fans with backward-skewed fan blades and a common operating point were designed with a 2D element blade method. Two approaches were adopted to modify the inflow conditions: first, the inflow turbulence intensity was increased by two different rectangular grids and second, the inflow velocity profile was changed to an asymmetric characteristic by two grids with a distinct bar stacking. An increase in the inflow turbulence intensity affects both tonal and broadband noise, whereas a non-uniform velocity profile at the inlet influences mainly tonal components. The magnitude of this effect is not the same for all fans but is dependent on the blade skew. The impact is greater for the forward-skewed fans than for the backward-skewed and thus directly linked to the fan blade geometry.

  11. Hemocompatibility of Axial Versus Centrifugal Pump Technology in Mechanical Circulatory Support Devices.

    Science.gov (United States)

    Schibilsky, David; Lenglinger, Matthias; Avci-Adali, Meltem; Haller, Christoph; Walker, Tobias; Wendel, Hans Peter; Schlensak, Christian

    2015-08-01

    The hemocompatible properties of rotary blood pumps commonly used in mechanical circulatory support (MCS) are widely unknown regarding specific biocompatibility profiles of different pump technologies. Therefore, we analyzed the hemocompatibility indicating markers of an axial flow and a magnetically levitated centrifugal device within an in vitro mock loop. The HeartMate II (HM II; n = 3) device and a CentriMag (CM; n = 3) adult pump were investigated in a human whole blood mock loop for 360 min using the MCS devices as a driving component. Blood samples were analyzed by enzyme-linked immunosorbent assay for markers of coagulation, complement system, and inflammatory response. There was a time-dependent activation of the coagulation (thrombin-antithrombin complexes [TAT]), complement (SC5b-9), and inflammation system (polymorphonuclear [PMN] elastase) in both groups. The mean value of TAT (CM: 4.0 μg/L vs. 29.4 μg/L, P centrifugal CM device showed significantly lower activation of coagulation and inflammation than that of the HM II axial flow pump. Both HM II and CM have demonstrated an acceptable hemocompatibility profile in patients. However, there is a great opportunity to gain a clinical benefit by developing techniques to lower the blood surface interaction within both pump technologies and a magnetically levitated centrifugal pump design might be superior.

  12. Radial breathing mode of carbon nanotubes subjected to axial pressure.

    Science.gov (United States)

    Lei, Xiao-Wen; Ni, Qing-Qing; Shi, Jin-Xing; Natsuki, Toshiaki

    2011-08-11

    In this paper, a theoretical analysis of the radial breathing mode (RBM) of carbon nanotubes (CNTs) subjected to axial pressure is presented based on an elastic continuum model. Single-walled carbon nanotubes (SWCNTs) are described as an individual elastic shell and double-walled carbon nanotubes (DWCNTs) are considered to be two shells coupled through the van der Waals force. The effects of axial pressure, wave numbers and nanotube diameter on the RBM frequency are investigated in detail. The validity of these theoretical results is confirmed through the comparison of the experiment, calculation and simulation. Our results show that the RBM frequency is linearly dependent on the axial pressure and is affected by the wave numbers. We concluded that RBM frequency can be used to characterize the axial pressure acting on both ends of a CNT.

  13. The geometrical theory of diffraction for axially symmetric reflectors

    DEFF Research Database (Denmark)

    Rusch, W.; Sørensen, O.

    1975-01-01

    The geometrical theory of diffraction (GTD) (cf. [1], for example) may be applied advantageously to many axially symmetric reflector antenna geometries. The material in this communication presents analytical, computational, and experimental results for commonly encountered reflector geometries...

  14. Time Domain Terahertz Axial Computed Tomography Non Destructive Evaluation Project

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to demonstrate key elements of feasibility for a high speed automated time domain terahertz computed axial tomography (TD-THz CT) non destructive...

  15. ANALYSIS OF PULSATILE BLOOD FLOW IN AXIALLY MOVING ARTERIES

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    In order to study motional properties of pulsatile blood flow in axially moving arteries, the authors derived some expressions of the pulsatile blood flow from the basic equations of motion for blood and vascular walls, including an axial blood velocity equation, a flow rate equation and a wall shear stress equation, which described not only the overall axial movement of the arteries but also the elastic properties of the vascular walls, discussed the effects of the arterial wall elasticity on the wall shear stress in coronary arteries in terms of these expressions, and analyzed changes of motional properties of pulsatile blood flow between an elastic arterial tube model and a rigid tube model. The results proved the inference by J.E. Moore Jr. et al. (1994) that the axial movement of arteries be as important in determining coronary artery hemodynamics as the elastic property of the vascular wall.

  16. Open-Axial-Field Magnet at ISR intersection I8

    CERN Multimedia

    1979-01-01

    This axial field spectrometer left the central collision region unobstructed in order to ease analysis of secondary particles emitted at large angle. The ISR circulating beams were passing through a hole in the magnet poles.

  17. Axial mesodermal dysplasia complex: a new case with parental consanguinity.

    Science.gov (United States)

    Mota, C R; Azevedo, M; Rocha, G; Manuela, F; Coelho, R; Lima, M R

    2000-01-01

    A female is described with axial mesodermal dysplasia complex (AMDC) born to a consanguineous couple. This is thought to be the first description of a patient with AMDC born to consanguineous parents.

  18. Axial Stringy System of the Kerr Spinning Particle

    CERN Document Server

    Burinskii, A

    2004-01-01

    The structure of classical spinning particle based on the Kerr-Newman black hole (BH) solution is investigated. For large angular momentum, $|a|>>m$, the BH horizons disappear exposing a naked ringlike source which is a circular relativistic string. It was shown recently that electromagnetic excitations of this string lead to the appearance of an extra axial stringy system which consists of two half-infinite strings of opposite chirality. In this paper we consider the relation of this stringy system to the Dirac equation. We also show that the axial strings are the Witten superconducting strings and describe their structure by the Higgs field model where the Higgs condensate is used to regularize axial singularity. We argue that this axial stringy system may play the role of a classical carrier of the wave function.

  19. Time Domain Terahertz Axial Computed Tomography Non Destructive Evaluation Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this Phase 2 project, we propose to develop, construct, and deliver to NASA a computed axial tomography time-domain terahertz (CT TD-THz) non destructive...

  20. Uniform Decay for Solutions of an Axially Moving Viscoelastic Beam

    Energy Technology Data Exchange (ETDEWEB)

    Kelleche, Abdelkarim, E-mail: kellecheabdelkarim@gmail.com [Université des Sciences et de la Technologie Houari Boumediene, Faculté des Mathématiques (Algeria); Tatar, Nasser-eddine, E-mail: tatarn@Kfupm.edu.sa [King Fahd University of Petroleum and Minerals, Department of Mathematics and Statistics (Saudi Arabia)

    2017-06-15

    The paper deals with an axially moving viscoelastic structure modeled as an Euler–Bernoulli beam. The aim is to suppress the transversal displacement (transversal vibrations) that occur during the axial motion of the beam. It is assumed that the beam is moving with a constant axial speed and it is subject to a nonlinear force at the right boundary. We prove that when the axial speed of the beam is smaller than a critical value, the dissipation produced by the viscoelastic material is sufficient to suppress the transversal vibrations. It is shown that the rate of decay of the energy depends on the kernel which arise in the viscoelastic term. We consider a general kernel and notice that solutions cannot decay faster than the kernel.

  1. Axial Anomaly in Lattice Abelian Gauge Theory in Arbitrary Dimensions

    CERN Document Server

    Fujiwara, T; Wu, K; Fujiwara, Takanori; Suzuki, Hiroshi; Wu, Ke

    1999-01-01

    Axial anomaly of lattice abelian gauge theory in hyper-cubic regular lattice in arbitrary even dimensions is investigated by applying the method of exterior differential calculus. The topological invariance, gauge invariance and locality of the axial anomaly determine the explicit form of the topological part. The anomaly is obtained up to a multiplicative constant for finite lattice spacing and can be interpreted as the Chern character of the abelian lattice gauge theory.

  2. Axially symmetric volume constrained anisotropic mean curvature flow

    CERN Document Server

    Palmer, Bennett

    2011-01-01

    We study the long time existence theory for a non local flow associated to a free boundary problem for a trapped non liquid drop. The drop has free boundary components on two horizontal plates and its free energy is anisotropic and axially symmetric. For axially symmetric initial surfaces with sufficiently large volume, we show that the flow exists for all time. Numerical simulations of the curvature flow are presented.

  3. Axial myopia in computed and magnetic resonance tomography

    Energy Technology Data Exchange (ETDEWEB)

    Beyer-Enke, S.A.; Goerich, J.; Gamroth, A.

    1987-08-01

    The case of a 44-year old woman suffering from amblyopia on the left eye with unilateral proptosis caused by axial (progressive) myopia is presented. The clinical and radiological findings were discussed in reference to the literature. The diagnosis was established by ruling out neoplastic, inflammatory or endocrine causes for the exophtalmos. CT and MR scans revealed an enlarged left globe without evidence of orbital masses. The findings were regarded as typical for the diagnosis at axial myopia.

  4. Watson's theorem and the $N\\Delta(1232)$ axial transition

    CERN Document Server

    Alvarez-Ruso, L; Nieves, J; Vacas, M J Vicente

    2016-01-01

    We present a new determination of the $N\\Delta$ axial form factors from neutrino induced pion production data. For this purpose, the model of Hernandez et al., Phys. Rev. D76, 033005 (2007) is improved by partially restoring unitarity. This is accomplished by imposing Watson's theorem on the dominant vector and axial multipoles. As a consequence, a larger $C_5^A(0)$, in good agreement with the prediction from the off-diagonal Goldberger-Treiman relation, is now obtained.

  5. Method to Measure Tone of Axial and Proximal Muscle

    OpenAIRE

    2011-01-01

    The control of tonic muscular activity remains poorly understood. While abnormal tone is commonly assessed clinically by measuring the passive resistance of relaxed limbs1, no systems are available to study tonic muscle control in a natural, active state of antigravity support. We have developed a device (Twister) to study tonic regulation of axial and proximal muscles during active postural maintenance (i.e. postural tone). Twister rotates axial body regions relative to each other about the ...

  6. Axially astigmatic surfaces: different types and their properties

    Science.gov (United States)

    Malacara-Doblado, Daniel; Malacara-Hernandez, Daniel; Garcia-Marquez, Jorge L.

    1996-12-01

    Axially astigmatic surfaces have different curvatures in orthogonal diameters. Toroidal and spherocylindrical optical surfaces are two mathematically different special cases of axially astigmatic surfaces as noted by Menchaca and Malacara (1986), but they are almost identical in the vicinity of the optical axis. The different between these two surfaces increases when the distance to the optical axis increases. We study the general properties of astigmatic surfaces and some special interesting cases.

  7. Axial blanket for 16NGF Angra 1 fuel type

    Energy Technology Data Exchange (ETDEWEB)

    Sadde, Luciano Martins; Faria, Eduardo Fernandes [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil)]. E-mails: sadde@inb.gov.br; faria@inb.gov.br; Sang-Keun You [Korea Nuclear Fuel Co. Ltd. (KNFC), Taejon (Korea, Republic of)]. E-mail: skyou@knfc.co.kr

    2007-07-01

    Angra-1, Kori-2 and Krsko are nuclear power plants with the same design. However, the fuel assemblies have some differences in design due to the countries strategies and the differences in the fabrication process. The 16NGF (16x16 Next Generation Fuel) was developed by INB, KNFC and Westinghouse in order to be used in these three nuclear power plants and the 'Axial Blanket' is one of the new features for the 16NGF design. The main purpose of the Axial Blanket Optimization study is to determine which axial blanket enrichment and length would provide the better fuel cycle cost benefit. All of the calculations were performed using Gadolinium as Burnable Absorber and solid pellets type for Axial Blanket. The results indicate 1.8 w/o U235 enrichment and 8 inches length as the best option of Axial Blanket from the fuel cycle cost benefit standpoint. The economy is about 1.8%. The difference in the reload cost in the range between 1.5 and 2.6 w/o U235 enrichment and for the 6 and 8 inches length is not so significant. Due that, from the Fq limit standpoint and also for longer cycle length requirements, a higher axial blanket enrichment (2.6 w/o) and shorter length (6 inches) is recommended. (author)

  8. Reactive control of subsonic axial fan noise in a duct.

    Science.gov (United States)

    Liu, Y; Choy, Y S; Huang, L; Cheng, L

    2014-10-01

    Suppressing the ducted fan noise at low frequencies without varying the flow capacity is still a technical challenge. This study examines a conceived device consisting of two tensioned membranes backed with cavities housing the axial fan for suppression of the sound radiation from the axial fan directly. The noise suppression is achieved by destructive interference between the sound fields from the axial fan of a dipole nature and sound radiation from the membrane via vibroacoustics coupling. A two-dimensional model with the flow effect is presented which allows the performance of the device to be explored analytically. The air flow influences the symmetrical behavior and excites the odd in vacuo mode response of the membrane due to kinematic coupling. Such an asymmetrical effect can be compromised with off-center alignment of the axial fan. Tension plays an important role to sustain the performance to revoke the deformation of the membrane during the axial fan operation. With the design of four appropriately tensioned membranes covered by a cylindrical cavity, the first and second blade passage frequencies of the axial fan can be reduced by at least 20 dB. The satisfactory agreement between experiment and theory demonstrates that its feasibility is practical.

  9. Slow axial drift in three-dimensional granular tumbler flow

    Science.gov (United States)

    Zaman, Zafir; D'Ortona, Umberto; Umbanhowar, Paul B.; Ottino, Julio M.; Lueptow, Richard M.

    2013-07-01

    Models of monodisperse particle flow in partially filled three-dimensional tumblers often assume that flow along the axis of rotation is negligible. We test this assumption, for spherical and double cone tumblers, using experiments and discrete element method simulations. Cross sections through the particle bed of a spherical tumbler show that, after a few rotations, a colored band of particles initially perpendicular to the axis of rotation deforms: particles near the surface drift toward the pole, while particles deeper in the flowing layer drift toward the equator. Tracking of mm-sized surface particles in tumblers with diameters of 8-14 cm shows particle axial displacements of one to two particle diameters, corresponding to axial drift that is 1-3% of the tumbler diameter, per pass through the flowing layer. The surface axial drift in both double cone and spherical tumblers is zero at the equator, increases moving away from the equator, and then decreases near the poles. Comparing results for the two tumbler geometries shows that wall slope causes axial drift, while drift speed increases with equatorial diameter. The dependence of axial drift on axial position for each tumbler geometry is similar when both are normalized by their respective maximum values.

  10. Axial scanning in confocal microscopy employing adaptive lenses (CAL).

    Science.gov (United States)

    Koukourakis, Nektarios; Finkeldey, Markus; Stürmer, Moritz; Leithold, Christoph; Gerhardt, Nils C; Hofmann, Martin R; Wallrabe, Ulrike; Czarske, Jürgen W; Fischer, Andreas

    2014-03-10

    In this paper we analyze the capability of adaptive lenses to replace mechanical axial scanning in confocal microscopy. The adaptive approach promises to achieve high scan rates in a rather simple implementation. This may open up new applications in biomedical imaging or surface analysis in micro- and nanoelectronics, where currently the axial scan rates and the flexibility at the scan process are the limiting factors. The results show that fast and adaptive axial scanning is possible using electrically tunable lenses but the performance degrades during the scan. This is due to defocus and spherical aberrations introduced to the system by tuning of the adaptive lens. These detune the observation plane away from the best focus which strongly deteriorates the axial resolution by a factor of ~2.4. Introducing balancing aberrations allows addressing these influences. The presented approach is based on the employment of a second adaptive lens, located in the detection path. It enables shifting the observation plane back to the best focus position and thus creating axial scans with homogeneous axial resolution. We present simulated and experimental proof-of-principle results.

  11. Relationship Between Retinal Vein Occlusion and Axial Length

    Directory of Open Access Journals (Sweden)

    San-Chang Tsai

    2003-09-01

    Full Text Available This study examined whether axial length is a local risk factor for central retinal vein occlusion (CRVO and branch retinal vein occlusion (BRVO. The study group consisted of 40 patients with unilateral CRVO and 77 patients with unilateral BRVO. The control group included 67 individuals who matched the study group patients in age, systemic hypertension, and diabetes mellitus status. The axial lengths of affected and fellow eyes of patients and controls were measured using A-scan ultrasonography. The axial length of affected eyes was statistically significantly shorter than that of unaffected eyes in the BRVO group (p < 0.05 but not in the CRVO group (p = 0.05. There were also statistically significant differences in axial length between control eyes and affected eyes in both the CRVO group (p < 0.05 and BRVO group (p < 0.05. Thus, shorter axial length could be a risk factor for developing CRVO and BRVO. The axial lengths of affected eyes in retinal vein occlusion patients tend to be shorter than those of unaffected eyes, especially in BRVO patients.

  12. Lipid Profile

    Science.gov (United States)

    ... AACC products and services. Advertising & Sponsorship: Policy | Opportunities Lipid Profile Share this page: Was this page helpful? Also ... as: Lipid Panel; Coronary Risk Panel Formal name: Lipid Profile Related tests: Cholesterol ; HDL Cholesterol ; LDL Cholesterol ; Triglycerides ; ...

  13. Data Profiling

    OpenAIRE

    Hladíková, Radka

    2010-01-01

    Title: Data Profiling Author: Radka Hladíková Department: Department of Software Engineering Supervisor: Ing. Vladimír Kyjonka Supervisor's e-mail address: Abstract: This thesis puts mind on problems with data quality and data profiling. This Work analyses and summarizes problems of data quality, data defects, process of data quality, data quality assessment and data profiling. The main topic is data profiling as a process of researching data available in existing...

  14. Axial and radial water transport and internal water storage in tropical forest canopy trees.

    Science.gov (United States)

    James, Shelley A; Meinzer, Frederick C; Goldstein, Guillermo; Woodruff, David; Jones, Timothy; Restom, Teresa; Mejia, Monica; Clearwater, Michael; Campanello, Paula

    2003-01-01

    Heat and stable isotope tracers were used to study axial and radial water transport in relation to sapwood anatomical characteristics and internal water storage in four canopy tree species of a seasonally dry tropical forest in Panama. Anatomical characteristics of the wood and radial profiles of sap flow were measured at the base, upper trunk, and crown of a single individual of Anacardium excelsum, Ficus insipida, Schefflera morototoni, and Cordia alliodora during two consecutive dry seasons. Vessel lumen diameter and vessel density did not exhibit a consistent trend axially from the base of the stem to the base of the crown. However, lumen diameter decreased sharply from the base of the crown to the terminal branches. The ratio of vessel lumen area to sapwood cross-sectional area was consistently higher at the base of the crown than at the base of the trunk in A. excelsum, F. insipida and C. alliodora, but no axial trend was apparent in S. morototoni. Radial profiles of the preceding wood anatomical characteristics varied according to species and the height at which the wood samples were obtained. Radial profiles of sap flux density measured with thermal dissipation sensors of variable length near the base of the crown were highly correlated with radial profiles of specific hydraulic conductivity (k(s)) calculated from xylem anatomical characteristics. The relationship between sap flux density and k(s) was species-independent. Deuterium oxide (D(2)O) injected into the base of the trunk of the four study trees was detected in the water transpired from the upper crown after only 1 day in the 26-m-tall C. alliodora tree, 2 days in the 28-m-tall F. insipida tree, 3 days in the 38-m-tall A. excelsum tree, and 5 days in the 22-m-tall S. morototoni tree. Radial transport of injected D(2)O was detected in A. excelsum, F. insipida and S. morototoni, but not C. alliodora. The rate of axial D(2)O transport, a surrogate for maximum sap velocity, was positively correlated

  15. Axial length as a risk factor to branch retinal vein occlusion

    NARCIS (Netherlands)

    Timmerman, EA; deLavalette, VWR; vandenBrom, HJB

    1997-01-01

    Purpose: To determine whether axial length is a factor in branch retinal vein occlusion. Methods: Axial length measurements in a group of 24 patients with a unilateral branch retinal Vein occlusion were compared with the axial length measurements in a control group. axial length measurements were ta

  16. Development and Testing of an Axial Halbach Magnetic Bearing

    Science.gov (United States)

    Eichenberg, Dennis J.; Gallo, Christopher A.; Thompson, William K.

    2006-01-01

    The NASA Glenn Research Center has developed and tested a revolutionary Axial Halbach Magnetic Bearing. The objective of this work is to develop a viable non-contact magnetic thrust bearing utilizing Halbach arrays for all-electric flight, and many other applications. This concept will help to reduce harmful emissions, reduce the Nation s dependence on fossil fuels and mitigate many of the concerns and limitations encountered in conventional axial bearings such as bearing wear, leaks, seals and friction loss. The Axial Halbach Magnetic Bearing is inherently stable and requires no active feedback control system or superconductivity as required in many magnetic bearing designs. The Axial Halbach Magnetic Bearing is useful for very high speed applications including turbines, instrumentation, medical systems, computer memory systems, and space power systems such as flywheels. Magnetic fields suspend and support a rotor assembly within a stator. Advanced technologies developed for particle accelerators, and currently under development for maglev trains and rocket launchers, served as the basis for this application. Experimental hardware was successfully designed and developed to validate the basic principles and analyses. The report concludes that the implementation of Axial Halbach Magnetic Bearings can provide significant improvements in rotational system performance and reliability.

  17. Method to measure tone of axial and proximal muscle.

    Science.gov (United States)

    Gurfinkel, Victor S; Cacciatore, Timothy W; Cordo, Paul J; Horak, Fay B

    2011-12-14

    The control of tonic muscular activity remains poorly understood. While abnormal tone is commonly assessed clinically by measuring the passive resistance of relaxed limbs, no systems are available to study tonic muscle control in a natural, active state of antigravity support. We have developed a device (Twister) to study tonic regulation of axial and proximal muscles during active postural maintenance (i.e. postural tone). Twister rotates axial body regions relative to each other about the vertical axis during stance, so as to twist the neck, trunk or hip regions. This twisting imposes length changes on axial muscles without changing the body's relationship to gravity. Because Twister does not provide postural support, tone must be regulated to counteract gravitational torques. We quantify this tonic regulation by the restive torque to twisting, which reflects the state of all muscles undergoing length changes, as well as by electromyography of relevant muscles. Because tone is characterized by long-lasting low-level muscle activity, tonic control is studied with slow movements that produce "tonic" changes in muscle length, without evoking fast "phasic" responses. Twister can be reconfigured to study various aspects of muscle tone, such as co-contraction, tonic modulation to postural changes, tonic interactions across body segments, as well as perceptual thresholds to slow axial rotation. Twister can also be used to provide a quantitative measurement of the effects of disease on axial and proximal postural tone and assess the efficacy of intervention.

  18. Design and Simulation of Axial Flow Maglev Blood Pump

    Directory of Open Access Journals (Sweden)

    Huachun Wu

    2011-03-01

    Full Text Available The axial flow maglev blood pump (AFMBP has become a global research focus and emphasis for artificial ventricular assist device, which has no mechanical contact, mechanical friction, compact structure and light weight, can effectively solve thrombus and hemolysis. Magnetic suspension and impeller is two of the important parts in the axial flow maglev blood pump, and their structure largely determines the blood pump performance. The research adopts electromagnetic and fluid finite element analysis, and puts forward a method to design the magnetic suspension and impeller of axial flow blood pump, which tacks into account the small volume of axial blood pump. The magnetic bearing’s characteristics are evaluated by electromagnetic finite element analysis. The Blades have been designed by calculating aerofoil bone line, and make simulation analysis for different thicken ways of blade by Fluent software, and make a conclusion that the blade thickened with certain rules has better characteristics in the same conditions. The results will provide some guidance for design of axial flow maglev blood pump, and establish theoretical basis for application of the implantable artificial heart pump.

  19. Light weakly coupled axial forces: models, constraints, and projections

    Science.gov (United States)

    Kahn, Yonatan; Krnjaic, Gordan; Mishra-Sharma, Siddharth; Tait, Tim M. P.

    2017-05-01

    We investigate the landscape of constraints on MeV-GeV scale, hidden U(1) forces with nonzero axial-vector couplings to Standard Model fermions. While the purely vector-coupled dark photon, which may arise from kinetic mixing, is a well-motivated scenario, several MeV-scale anomalies motivate a theory with axial couplings which can be UV-completed consistent with Standard Model gauge invariance. Moreover, existing constraints on dark photons depend on products of various combinations of axial and vector couplings, making it difficult to isolate the effects of axial couplings for particular flavors of SM fermions. We present a representative renormalizable, UV-complete model of a dark photon with adjustable axial and vector couplings, discuss its general features, and show how some UV constraints may be relaxed in a model with nonrenormalizable Yukawa couplings at the expense of fine-tuning. We survey the existing parameter space and the projected reach of planned experiments, briefly commenting on the relevance of the allowed parameter space to low-energy anomalies in π0 and 8Be∗ decay.

  20. An Axial Vector Photon in a Mirror World

    CERN Document Server

    Sharafiddinov, Rasulkhozha S

    2015-01-01

    The unity of symmetry laws emphasizes, in the case of a mirror CP-even Dirac Lagrangian, the ideas of the left- and right-handed axial-vector photons referring to long- and short-lived bosons of true neutrality, respectively. Such a difference in lifetimes expresses the unidenticality of masses, energies and momenta of axial-vector photons of the different components. They define the unified field theory equation of C-odd particles with an integral spin. Together with a new equation of a theory of truly neutral particles with the half-integral spin, the latter reflects the availability in their nature of the second type of the local axial-vector gauge transformation responsible for origination in the Lagrangian of C-oddity of an interaction Newton component giving an axial-vector mass to all the interacting particles and fields. The mirror axial-vector mass, energy and momentum operators constitute a CP-invariant equation of quantum mechanics, confirming that each of them can individually influence on matter ...