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Sample records for high burnup plutonium-aluminum

  1. High Burnup Fuel Behavior Modeling

    SciTech Connect

    Jahingir, M.; Rand, R.; Stachowski, R.; Miles, B.; Kusagaya, K.

    2007-07-01

    This paper discusses the development and qualification of the PRIME03 code to address high burnup mechanisms and to improve uranium utilization in current and new reactor designs. Materials properties and behavioral models have been updated from previous thermal-mechanical codes to reflect the effects of burnup on fuel pellet thermal conductivity, Zircaloy creep, fuel pellet relocation, and fission gas release. These new models are based on results of in-pool and post irradiation examination (PIE) of commercial boiling water reactor (BWR) fuel rods at high burnup and results from international experimental programs. The new models incorporated into PRIME03 also address specific high burnup effects associated with formation of pellet rim porosity at high exposure. The PRIME03 code is qualified by comparison of predicted and measured fuel performance parameters for a large number of high, low, and moderate burnup test and commercial reactor rod. The extensive experimental qualification of the PRIME03 prediction capabilities confirms that it is a reliable best-estimate predictor of fuel rod thermal-mechanical performance over a wide range of design and operating conditions. (authors)

  2. High burnup effects in WWER fuel rods

    SciTech Connect

    Smirnov, V.; Smirnov, A.

    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.

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

    SciTech Connect

    Diamond, D.J.; Neymoith, L.; Kohut, P.

    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.

  4. Effects of Burnup and Temperature Distributions to CANDLE Burnup of Block-Type High Temperature Gas Cooled Reactor

    SciTech Connect

    Yasunori Ohoka; Ismile; Hiroshi Sekimoto

    2004-07-01

    The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top or from top to bottom of the core and without any change in their shapes. It can be applied easily to the block-type high temperature gas cooled reactor using an appropriate burnable poison mixed with uranium oxide fuel. In the present study, the burnup distribution and the temperature distribution in the core are investigated and their effects on the CANDLE burnup core characteristics are studied. In this study, the natural gadolinium is used as the burnable poison. With the fuel enrichment of 15%, the natural gadolinium concentration of 3.0% and the fuel pin pitch of 6.6 cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half width of power density distribution of 1.5 m for uniform group constant case at 900 K. When the effect of nuclide change by burnup is considered, the burning region speed becomes 25 cm/year and the axial half-width of power density distribution becomes 1.25 m. When the temperature distributions effect is considered, the effects on the core characteristics are smaller than the burnup distribution effect. The maximum fuel temperature of the parallel flow case is higher than the counter flow case. (authors)

  5. High-burnup core design using minor actinide-containing metal fuel

    SciTech Connect

    Ohta, Hirokazu; Ogata, Takanari; Obara, T.

    2013-07-01

    A neutronic design study of metal fuel fast reactor (FR) cores is conducted on the basis of an innovative fuel design concept to achieve an extremely high burnup and realize an efficient fuel cycle system. Since it is expected that the burnup reactivity swing will become extremely large in an unprecedented high burnup core, minor actinides (MAs) from light water reactors (LWRs) are added to fresh fuel to improve the core internal conversion. Core neutronic analysis revealed that high burnups of about 200 MWd/kg for a small-scale core and about 300 MWd/kg for a large-scale core can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup. An actinide burnup analysis has shown that the MA consumption ratio is improved to about 60% and that the accumulated MAs originating from LWRs can be efficiently consumed by the high-burnup metal fuel FR. (authors)

  6. Calculations on fission gas behaviour in the high burnup structure

    NASA Astrophysics Data System (ADS)

    Blair, P.; Romano, A.; Hellwig, Ch.; Chawla, R.

    2006-05-01

    The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing a one-dimensional, mass balance model and apply it to LWR UO 2 fuel at the moderate temperatures found in the rim region. We examine the quantity of gas remaining in the HBS fuel matrix at steady state and compare it with experimental values. We find that the current model reproduces the 0.2 wt% observed xenon concentration under certain conditions, viz. fast grain boundary diffusion and an effective volume diffusion coefficient. A sensitivity analysis is also conducted for the model parameters, the relative importance for which is not well established a priori.

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

    SciTech Connect

    Wang, Jy-An John; Wang, Hong

    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).

  8. Assessment of reactivity transient experiments with high burnup fuel

    SciTech Connect

    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.

  9. Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel

    SciTech Connect

    Melissa C. Teague; Brian P. Gorman; Steven L. Hayes; Douglas L. Porter; Jeffrey King

    2013-10-01

    High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

  10. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-12

    ... Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research and... notice of request for public comment on its draft test plan for the High Burnup Dry Storage Cask Research... development throughout the execution of the High Burnup Dry Storage Cask Research and Development Project....

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

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao; Bevard, Bruce Balkcom; Howard, Rob L; Scaglione, John M

    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.

  12. 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

    SciTech Connect

    GrandJean, C.; Cauvin, R.; Lebuffe, C.

    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.

  13. High Burnup Effects Program A State-of-the-Technology Assessment

    SciTech Connect

    Rising, K. H.; Bradley, E. R.; Williford, R. E.; Freshley, M D.

    1982-06-01

    Various analytical models and empirical correlations describing the fission gas release phenomenon were examined. An evaluation was made of the current pertinent experimental data on the subject of high burnup fission gas release. Data reported by individual investigators were compared and evaluated in relation to their applicability to the content and scope of the High Burnup Effects Program. These evaluations then form the bases for defining the data needs, and the selection of variables to be studied in this program.

  14. High burn-up structure of U(Mo) dispersion fuel

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Van Renterghem, W.; Van den Berghe, S.

    2016-08-01

    The evolution of the high burn-up structure (HBS) in U(Mo) fuel irradiated up to a burn-up of ∼70% 235U or ∼5 × 1021 f/cm3 or ∼120 GWd/tHM is described and compared to the observation made on LWR fuel. Scanning and transmission electron microscopy was performed on several samples having different burn-ups in order to get a better understanding of the mechanisms leading to the high burn-up structure formation. Even though there are some substantial differences between the irradiation of ceramic and U(Mo) alloy fuels (crystal structure, enrichment, irradiation temperature …), it was found that in both fuels recrystallization initiates at the same threshold and progresses in a similar way with increasing fission density. In case of U(Mo), recrystallization leads to accelerated swelling of the fuel which could result in instability of the fuel plate.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide

    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.

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

    SciTech Connect

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

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Wiesenack, W.

    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.

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

    SciTech Connect

    Ishijima, Kiyomi; Fuketa, Toyoshi

    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.

  2. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    SciTech Connect

    Bai, Xian -Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2015-12-22

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctly predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. Furthermore, an analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.

  3. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    DOE PAGESBeta

    Bai, Xian -Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2015-12-22

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctlymore » predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. Furthermore, an analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.« less

  4. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    NASA Astrophysics Data System (ADS)

    Bai, Xian-Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2016-03-01

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctly predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. An analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.

  5. Nuclide analysis in high burnup fuel samples irradiated in Vandellós 2

    NASA Astrophysics Data System (ADS)

    Zwicky, H. U.; Low, J.; Granfors, M.; Alejano, C.; Conde, J. M.; Casado, C.; Sabater, J.; Lloret, M.; Quecedo, M.; Gago, J. A.

    2010-07-01

    In the framework of a high burnup fuel demonstration programme, rods with an enrichment of 4.5% 235U were operated to a rod average burnup of about 70 MWd/kgU in the Spanish Vandellós 2 pressurised water reactor. The rods were sent to hot cells and used for different research projects. This paper describes the isotopic composition measurements performed on samples of those rods, and the analysis of the measurement results based on comparison against calculated values. The fraction and composition of fission gases released to the rod free volume was determined for two of the rods. About 8% of Kr and Xe produced by fission were released. From the isotopic composition of the gases, it could be concluded that the gases were not preferentially released from the peripheral part of the fuel column. Local burnup and isotopic content of gamma emitting nuclides were determined by quantitatively evaluating axial gamma scans of the full rods. Nine samples were cut at different axial levels from three of the rods and analysed in two campaigns. More than 50 isotopes of 16 different elements were assessed, most of them by Inductively Coupled Plasma Mass Spectrometry after separation with High Performance Liquid Chromatography. In general, these over 400 data points gave a consistent picture of the isotopic content of irradiated fuel as a function of burnup. Only in a few cases, the analysis provided unexpected results that seem to be wrong, in most cases due to unidentified reasons. Sample burnup analysis was performed by comparing experimental isotopic abundances of uranium and plutonium composition as well as neodymium isotopic concentrations with corresponding CASMO based data. The results were in agreement with values derived independently from gamma scanning and from core design data and plant operating records. Measured isotope abundances were finally assessed using the industry standard SAS2H sequence of the SCALE code system. This exercise showed good agreement between

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

    SciTech Connect

    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.

  7. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    SciTech Connect

    Chung, H.M. )

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 {mu}m in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307{degree}C rather than the normal 288{degree}C, a relatively thick (50 to 70 {mu}m) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs.

  8. Test plan for high-burnup fuel cladding behavior under loss-of- coolant accident conditions

    SciTech Connect

    Chung, H.M.; Neimark, L.A.; Kassner, T.F.

    1996-10-01

    Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding, and as result, mechanical properties of high-burnup fuels are degraded significantly. This may influence the current fuel cladding failure limits for loss-of- coolant-accident (LOCA) situations, which are based on fuel cladding behavior for zero burnup. To avoid cladding fragmentation and fuel dispersal during a LOCA, 10 CFR 50.46 requires that peak cladding temperature shall not exceed 1204 degrees C (2200 degrees F) and that total oxidation of the fuel cladding nowhere exceeds 0.17 times total cladding thickness before oxidation. Because of the concern, a new experimental program to investigate high-burnup fuel cladding behavior under LOCA situations has been initiated under the sponsorship of the U.S. Nuclear Regulatory Commission. A hot-cell test plan to investigate single-rod behavior under simulated LOCA conditions is described in this paper. In the meantime, industry fuel design and operating conditions are expected to undergo further changes as more advanced cladding materials are developed. Under these circumstances, mechanical properties of high-burnup fuel cladding require further investigation so that results from studies on LOCA, reactivity- initiated-accident (RIA), operational transient, and power-ramping situations, can be extrapolated to modified or advanced cladding materials and altered irradiation conditions without repeating major integral experiments in test reactors. To provide the applicable data base and mechanistic understanding, tests will be conducted to determine dynamic and static fracture toughness and tensile properties. Background and rationale for selecting the specific mechanical properties tests are also described.

  9. Mechanistic model for the fragmentation of the high-burnup structure during LOCA

    NASA Astrophysics Data System (ADS)

    Kulacsy, Katalin

    2015-11-01

    A model was developed to account for the fragmentation of the high-burnup structure in fuel pellets during a loss-of-coolant accident. The basic assumptions of the model are that the pores have reached the dislocation punching overpressure during base irradiation and that the increasing overpressure due to rising temperature causes fuel fracturing. The distribution of the pore sizes is taken into account, not only the mean value. The model predicts the existence of a minimum pore size in normal operation, the decrease of threshold burnup for fuel fragmentation when the last-cycle power rating of the rod is high and a reduced fragmentation when a strong PCMI restraint is present during the transient. It can reproduce experimental results in terms of temperature range for fragmentation.

  10. Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor

    NASA Astrophysics Data System (ADS)

    Yan, Wei-Hua; Zhang, Li-Guo; Zhang, Yan; Zhang, Zhao; Xiao, Zhi-Gang

    2013-11-01

    Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanium (HPGe) gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the 137Cs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% (1σ). The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors.

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

    SciTech Connect

    Stockman, Christine T.; Alsaed, Halim A.; Bryan, Charles R.

    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.

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

    SciTech Connect

    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.

  13. Effect of high burn-up and MOX fuel on reprocessing, vitrification and disposal of PWR and BWR spent fuels based on accurate burn-up calculation

    SciTech Connect

    Yoshikawa, T.; Iwasaki, T.; Wada, K.; Suyama, K.

    2006-07-01

    To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated bum-up calculation code systems combined with the bum-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel. (authors)

  14. Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel

    NASA Astrophysics Data System (ADS)

    Teague, Melissa; Tonks, Michael; Novascone, Stephen; Hayes, Steven

    2014-01-01

    Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON [1] fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez-Lucuta model was favorable.

  15. Interaction of dislocations in UO2 during high burn-up structure formation

    NASA Astrophysics Data System (ADS)

    Baranov, V. G.; Lunev, A. V.; Tenishev, A. V.; Khlunov, A. V.

    2014-01-01

    Dislocation dynamics is used to investigate the distribution of dislocations in oxide nuclear fuel under irradiation using the values of dislocation density from experiments. A model is constructed to account for the effects of irradiation on dislocation movement and for the brittle behavior of the material. Results show that the ground state of interacting dislocations in UO2 during irradiation is a periodic structure with spacing between walls equal to 100-300 nm at experimental dislocation densities. These regions adorned by dislocation walls are called sub-grains and represent the result of polygonization. The threshold of polygonization is shown to depend on the fluctuations of the stress field produced by interaction of many dislocations. These fluctuations reach a critical value when a critical dislocation density is reached (˜4 × 1014 m-2). The calculated value matches experimental data on dislocation density measurement of irradiated uranium dioxide at burn-up corresponding to the formation of high burn-up structure.

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

    SciTech Connect

    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.

  17. Microstructural Modeling of Thermal Conductivity of High Burn-up Mixed Oxide Fuel

    SciTech Connect

    Melissa Teague; Michael Tonks; Stephen Novascone; Steven Hayes

    2014-01-01

    Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez–Lucuta model was favorable.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Ilas, Germina; Gauld, Ian C

    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.

  20. Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA

    SciTech Connect

    Wenfeng Liu; Kazimi, Mujid S.

    2006-07-01

    This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0--56 MW d/kg), peak fuel enthalpy deposit (70--190 cal/g), degree of subcooling (0--80 deg. C), and extent of oxidation (0--25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface

  1. R and D of Oxide Dispersion Strengthening Steels for High Burn-up Fuel Claddings

    SciTech Connect

    Kimura, A.; Cho, H.S.; Lee, J.S.; Kasada, R.; Ukai, S.; Fujiwara, M.

    2004-07-01

    Research and development of fuel clad materials for high burn-up operation of light water reactor and super critical water reactor (SCPWR) will be shown with focusing on the effort to overcome the requirements of material performance as the fuel clad. Oxide dispersion strengthening (ODS) steels are well known as a high temperature structural material. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening without accompanying loss of ductility. High Cr ODS steels whose chromium concentration was in the range from 15 to 19 wt% showed high resistance to corrosion in supercritical pressurized water (SCPW). As for the susceptibility to hydrogen embrittlement of ODS steels, the critical hydrogen concentration required to hydrogen embrittlement is ranging 10{approx}12 wppm that is approximately one order of magnitude higher value than that of 9Cr reduced activation ferritic (RAF) steel. In the ODS steels, the fraction of helium desorption by bubble migration mechanism was smaller than that in the RAF steel, indicating that the ODS steels are also resistant to helium He bubble-induced embrittlement. Finally, it is demonstrated that the ODS steels are very promising for the fuel clad material for high burn-up operation of water-cooling reactors. (authors)

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

    SciTech Connect

    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

  3. Thermal Behavior of Advanced UO{sub 2} Fuel at High Burnup

    SciTech Connect

    Muller, E.; Lambert, T.; Silberstein, K.; Therache, B.

    2007-07-01

    To improve the fuel performance, advanced UO{sub 2} products are developed to reduce significantly Pellet-Cladding Interaction and Fission Gas Release to increase high burnup safety margins on Light Water Reactors. To achieve the expected improvements, doping elements are currently used, to produce large grain viscoplastic UO{sub 2} fuel microstructures. In that scope, AREVA NP is conducting the qualification of a new UO{sub 2} fuel pellet obtained by optimum chromium oxide doping. To assess the fuel thermal performance, especially the fuel conductivity degradation with increasing burnup and also the kinetics of fission gas release under transient operating conditions, an instrumented in-pile experiment, called REMORA, has been developed by the CEA. One segment base irradiated for five cycles in a French EDF commercial PWR ({approx} 62 GWd/tM) was consequently re-instrumented with a fuel centerline thermocouple and an advanced pressure sensor. The design of this specific sensor is based on the counter-pressure principle and avoids any drift phenomenon due to nuclear irradiation. This rodlet was then irradiated in the GRIFFONOS rig of the Osiris experimental reactor at CEA Saclay. This device, located in the periphery of the core, is designed to perform test under conditions close to those prevailing in French PWR reactor. Power variations are carried out by translating the device relatively to the core. Self - powered neutron detectors are positioned in the loop in order to monitor the power the whole time of the irradiation. The re-irradiation of the REMORA experiment consisted of a stepped ramp to power in order to point out a potential degradation of the fuel thermal conductivity with increasing burnup. During the first part of the irradiation, most of the measurements were performed at low power in order to take into account the irradiation effects on UO{sub 2} thermal conductivity at high burnup in low range of temperature. The second part of the irradiation

  4. EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

    SciTech Connect

    Teague, Melissa C.; Gorman, Brian P.; Miller, Brandon D.; 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 approximately 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 approximately 2.5x 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 approximately 25 um 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.

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

    NASA Astrophysics Data System (ADS)

    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

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

    SciTech Connect

    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.

  7. Progress in the Research Programs to Elucidate Axial Cracking Fuel Failure at High Burnup

    SciTech Connect

    Ogata, Keizo; Aomi, Masaki; Baba, Toshikazu; Kamimura, Katsuichiro; Etoh, Yoshinori; Ito, Kunio; Kido, Toshiya; Teshima, Hideyuki

    2007-07-01

    A fuel failure with an axial crack starting outside the cladding and penetrating inwards was experienced by high burnup BWR fuel rods in power ramp test. On the other hand, no fuel failure caused by power ramp test has been currently reported on PWR fuel rods at burnups higher than 50 GWd/t. Extensive research programs regarding hydrogen behaviors and mechanical performances on irradiated BWR and PWR fuel claddings have been carried out to clarify the mechanism of the axial cracking and to quantify the conditions to cause fuel failure. Hydrogen solid solubility measurement on irradiated Zircaloy-2 materials showed almost comparable results to those on unirradiated ones. Hydride re-distribution and re-orientation behaviors were tested by heating irradiated BWR claddings with Zr-liner under the conditions of applied radial heat flux (temperature gradient) and circumferential stress. Mechanical performances of BWR claddings were evaluated mainly by the internal pressurizing tests. Internal pressurization tests applying various pressurizing sequences, e.g. stepwise increase in pressure with holding intervals, were also conducted to simulate crack propagation behaviors. Some specimens demonstrated characteristic fracture surfaces similar to those observed on the failed fuel rods after the power ramp. Mechanical performances of irradiated PWR claddings were tested at temperatures of 573 to 723 K. Metallographic examination after tensile tests revealed a large number of incipient cracks within the region of cladding outer rim where a concentrated hydride layer (hydride rim) has been formed during irradiation. Crack propagation test using an expanding mandrel device demonstrated the crack propagation at 573 K but no propagation at 658 K. (authors)

  8. Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

    NASA Astrophysics Data System (ADS)

    Billone, M. C.; Burtseva, T. A.; Einziger, R. E.

    2013-02-01

    Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with <2% offset strain prior to >50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520-620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350-650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for

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

    SciTech Connect

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

    2014-04-01

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

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

    SciTech Connect

    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.

  11. A Modal Expansion Equilibrium Cycle Perturbation Method for Optimizing High Burnup Fast Reactors

    NASA Astrophysics Data System (ADS)

    Touran, Nicholas W.

    This dissertation develops a simulation tool capable of optimizing advanced nuclear reactors considering the multiobjective nature of their design. An Enhanced Equilibrium Cycle (EEC) method based on the classic equilibrium method is developed to evaluate the response of the equilibrium cycle to changes in the core design. Advances are made in the consideration of burnup-dependent cross sections and dynamic fuel performance (fission gas release, fuel growth, and bond squeeze-out) to allow accuracy in high-burnup reactors such as the Traveling Wave Reactor. EEC is accelerated for design changes near a reference state through a new modal expansion perturbation method that expands arbitrary flux perturbations on a basis of λ-eigenmodes. A code is developed to solve the 3-D, multigroup diffusion equation with an Arnoldi-based solver that determines hundreds of the reference flux harmonics and later uses these harmonics to determine expansion coefficients required to approximate the perturbed flux. The harmonics are only required for the reference state, and many substantial and localized perturbations from this state are shown to be well-approximated with efficient expressions after the reference calculation is performed. The modal expansion method is coupled to EEC to produce the later-in-time response of each design perturbation. Because the code determines the perturbed flux explicitly, a wide variety of core performance metrics may be monitored by working within a recently-developed data management system called the ARMI. Through ARMI, the response of each design perturbation may be evaluated not only for the flux and reactivity, but also for reactivity coefficients, thermal hydraulics parameters, economics, and transient performance. Considering the parameters available, an automated optimization framework is designed and implemented. A non-parametric surrogate model using the Alternating Conditional Expectation (ACE) algorithm is trained with many design

  12. Impacts of a high-burnup spent fuel on a geological disposal system design

    SciTech Connect

    Cho, D.K.; Lee, Y.; Lee, J.Y.; Choi, H.J.; Choi, J.W.

    2007-07-01

    The influence of a burnup increase of a spent nuclear fuel on a deep geological disposal system was evaluated in this study. First, the impact of a burnup increase on each aspect related to thermal and nuclear safety concerns was quantified. And then, the tunnel length, excavation volume, and the raw materials for a cast insert, copper, bentonite, and backfill needed to constitute a disposal system were comprehensively analyzed based on the spent fuel inventory to generate 1 Terawatt-year (TWa), to establish the overall effects and consequences on a geological disposal. As a result, impact of a burnup increase on the criticality safety and radiation shielding was shown to be negligible. The disposal area, however, is considerably affected because of a higher thermal load. And, it is reasonable to use a canister such as the Korean Reference Disposal Canister (KDC-1) containing 4 spent fuels up to 50 GWD/MtU, and to use a canister containing 3 spent fuels beyond 50 GWD/MtU. Although a considerable increased, 33 % in the tunnel length and 30 % in the excavation volume, was observed as the burnup increases from 50 to 60 GWD/MtU, because a decrease in the canister needs can offset an increase in the excavation volume, it can be concluded that a burnup increase of a spent fuel is not a critical concern for a geological disposal of a spent fuel. (authors)

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Asmolov, V.; Yegorova, L.

    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.

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

    SciTech Connect

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

    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.

  17. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    SciTech Connect

    Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle; Wagner, John C.

    2013-07-01

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup

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

    SciTech Connect

    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

  19. Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

    NASA Astrophysics Data System (ADS)

    Laux, D.; Baron, D.; Despaux, G.; Kellerbauer, A. I.; Kinoshita, M.

    2012-01-01

    We report the measurement of elastic constants of non-irradiated UO 2, SIMFUEL (simulated spent fuel: UO 2 with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young's modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.

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

    SciTech Connect

    Salay, Michael; Gauntt, Randall O.; Lee, Richard Y.; 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.

  1. Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel

    SciTech Connect

    Gauld, I. C.; Ryman, J. C.

    2000-12-11

    This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance. The importance is investigated as a function of increasing burnup to assist in identifying the key changes in spent fuel characteristics between conventional- and extended-burnup regimes. Studies involving both pressurized water-reactor (PWR) fuel assemblies and boiling-water-reactor (BWR) assemblies are included. This study is seen to be a necessary first step in identifying the high-burnup spent fuel characteristics that may adversely affect the accuracy of current computational methods and data, assess the potential impact on previous guidance on isotopic source terms and decay-heat values, and thus help identify areas for methods and data improvement. Finally, several recommendations on the direction of possible future code validation efforts for high-burnup spent fuel predictions are presented.

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

    NASA Astrophysics Data System (ADS)

    Valance, Stéphane; Bertsch, Johannes

    2015-09-01

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

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

    SciTech Connect

    Waeckel, N.; Cauvin, R.; Lebuffe, C.

    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.

  4. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  5. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

    SciTech Connect

    Uwaba, Tomoyuki; Ito, Masahiro; Mizuno, Tomoyasu; Katsuyama, Kozo; Makenas, Bruce J.; Wootan, David W.; Carmack, Jon

    2011-06-16

    The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39E26 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

  6. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

    SciTech Connect

    Tomoyuki Uwaba; Masahiro Ito; Kozo Katsuyama; Bruce J. Makenas; David W. Wootan; Jon Carmack

    2011-05-01

    The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 1026 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

  7. Alteration Behavior of High Burnup Spent Fuel in Salt Brine Under Hydrogen Overpressure and in Presence of Bromide

    SciTech Connect

    Loida, Andreas; Metz, Volker; Kienzler, Bernhard

    2007-07-01

    Recent studies have shown that in the presence of H2 overpressure, which forms due to the corrosion of the Fe based container, the dissolution rate of the spent fuel matrix is slowed down by a factor of about 10, associated with a distinct decrease of concentrations of important radionuclides. However, in a natural salt environment as well as in geological formations with chloride rich groundwater the presence of radiation chemically active impurities such as bromide must be taken in consideration. Bromide is known to react with {beta}/{gamma} radiolysis products, thus counteracting the protective H{sub 2} effect. In the present experiments using high burnup spent fuel, it is observed that during 212 days the matrix dissolution rate was enhanced by a factor of about 10 in the presence of up to 10{sup -3} M bromide and 3.2 bar H{sub 2} overpressure. However, concentrations of matrix bound actinides were found at the same level or below as found under identical conditions, but in the absence of bromide. In the long-term it is expected that the effect of bromide becomes less important, because the decrease of {beta}/{gamma}-activity results in a decrease of oxidative radicals, which react with bromide, while a-activity will dominate the radiation field. (authors)

  8. Relationship between changes in the crystal lattice strain and thermal conductivity of high burnup UO 2 pellets

    NASA Astrophysics Data System (ADS)

    Amaya, Masaki; Nakamura, Jinichi; Fuketa, Toyoshi; Kosaka, Yuji

    2010-01-01

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

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

    NASA Astrophysics Data System (ADS)

    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. Separation of metallic residues from the dissolution of a high-burnup BWR fuel using nitrogen trifluoride

    SciTech Connect

    McNamara, Bruce K.; Buck, Edgar C.; Soderquist, Chuck Z.; Smith, Frances N.; Mausolf, Edward J.; Scheele, Randall D.

    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.

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

    NASA Astrophysics Data System (ADS)

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

    2012-01-01

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

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

    NASA Astrophysics Data System (ADS)

    Baranov, V. G.; Lunev, A. V.; Reutov, V. F.; Tenishev, A. V.; Isaenkova, M. G.; Khlunov, A. V.

    2014-09-01

    Experiments in IC-100 and U-400 cyclotrons were conducted with SIMFUEL pellets (11.47 wt.% of fission products simulators) to reproduce some aspects of the long-term irradiation conditions in epithermal reactors. Pellets were irradiated with Xe16+, Xe24+ and He+ at energies ranging from 20 keV (He+) to 320 keV (Xe16+) and 1-90 MeV (Xe24+). Some samples were subsequently annealed to obtain larger grain sizes and to study defects recovery. The major microstructural changes consisted in grain sub-division observed on SEM and AFM images and change in composition registered by EPMA (pellets irradiated with 1-90 MeV Xe24+ ions at fluence of 5 × 1015 cm-2). Lattice distortion and increase in dislocation density is also noted according to X-ray data. At low energies and high fluences formation of bubbles (20 keV He+ at 5.5 × 1017 cm-2) was observed. Grain sub-division exhibits full coverage of the grain body and preservation of former grain boundaries. The size of sub-grains depends on local dislocation density and changes from 200 nm to 400 nm along the irradiated surface. Beneath it the size ranges from 150 to 600 nm. Sub-grains are not observed in samples irradiated by low-energy ions even at high dislocation densities.

  13. FRAPTRAN Predictability of High Burnup Advanced Fuel Performance: Analysis of the CABRI CIP0-1 and CIP0-2 Experiments

    SciTech Connect

    Del Barrio, M.T.; Herranz, L.E.

    2007-07-01

    Adequacy of analytical tools to estimate advanced high burnup fuel during a power pulse need to be soundly proven. Most of models in codes dealing with transient are extrapolations of those developed for lower irradiations. In addition, lack of open information prevents often a proper account of mechanical properties of new advanced cladding material. These circumstances make experimental programs on high burnup fuel performance an indispensable tool to enhance safety codes predictability through building up sound databases on which models can be extended or developed and on which suitable code performance can be proven. The experiments CIP0-1 and CIP0-2, carried out on 2002 in the CABRI reactor, can be seen as reference tests to investigate high burnup fuel response to RIA transients. Fuel rods of up to 75 GWd/tU (average rod burnup) encapsulated in advanced cladding materials (ZIRLO and M5) were submitted to power pulses of about 30 ms of half maximum width that injected 90-100 cal/g after 1.2 s. None of the rodlets failed during the experiments, but they underwent deformation that was experimentally determined. The FRAPTRAN code has been used for the analysis of these RIA tests. The fuel rod characterization necessary for FRAPTRAN at the end of the base irradiation, prior to the transient, was provided by FRAPCON-3. An investigation of major deviations of fuel rod characterization at the end of the base irradiation has highlighted that thermal uncertainties could result in outstanding discrepancies in FGR estimates. Transient comparison with the available data shows that FRAPTRAN presents a relatively good agreement in permanent clad hoop strain and overestimates significantly the axial elongation of the cladding. The potential effect of approximations made in describing the cladding mechanical behavior, the fuel-to-clad relative movement and the pre-transient gap width, have been all discussed. Given existing uncertainties, a conclusive statement could not be

  14. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

    SciTech Connect

    Wagner, J.C.; DeHart, M.D.

    2000-03-01

    This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.

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

    SciTech Connect

    Medvedev, Pavel G; Ozaltun, Hakan; Robinson, Adam Brady; Rabin, Barry H

    2014-04-01

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

  16. JAEA Studies on High Burnup Fuel Behaviors during Reactivity-Initiated Accident and Loss-of-Coolant Accident

    SciTech Connect

    Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nagase, Fumihisa; Suzuki, Motoe

    2007-07-01

    The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA program is comprised of reactivity-initiated accident (RIA) studies including pulse-irradiation experiments in the NSRR and cladding mechanical tests, loss-of-coolant accident (LOCA) tests including integral thermal shock test and oxidation rate measurement, development and verification of computer codes FEMAXI-6 and RANNS, and so on. In addition to an overview of the fuel safety research at JAEA, most recent progresses in the RIA and LOCA tests programs and the codes development are described and discussed in the paper. (authors)

  17. Selenium redox speciation and coordination in high-burnup UO2 fuel: Consequences for the release of 79Se in a deep underground repository

    NASA Astrophysics Data System (ADS)

    Curti, Enzo; Froideval-Zumbiehl, Annick; Günther-Leopold, Ines; Martin, Matthias; Bullemer, Andrej; Linder, Hanspeter; Borca, Camelia N.; Grolimund, Daniel

    2014-10-01

    The chemical state of Se in high-burnup UO2 spent nuclear fuel (SNF) was investigated by micro X-ray absorption near-edge structure (μ-XANES) spectroscopy. The data were first evaluated in terms of linear combination fits of reference compounds. In a second step, the XANES data were fitted to theoretical (ab initio) spectra via geometrical optimization of atomic clusters around a central Se atom, using the FDMNES and FitIt software packages. Best fits were obtained assuming substitution of Se in occupied or vacant oxygen sites within the UO2 lattice, with 20-25% local expansion. Based on these results and chemical arguments we argue that the long-lived fission product 79Se may be stabilized to sparingly soluble Se(-II) in SNF. This would explain the failure to detect dissolved Se in SNF leaching experiments.

  18. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    NASA Astrophysics Data System (ADS)

    Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji

    2011-09-01

    The C3M irradiation test, which was conducted in the experimental fast reactor, "Joyo", demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, "Monju". The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  19. Local Burn-Up Effects in the NBSR Fuel Element

    SciTech Connect

    Brown N. R.; Hanson A.; Diamond, D.

    2013-01-31

    This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peaking relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.

  20. CANDLE: The New Burnup Strategy

    SciTech Connect

    Sekimoto, Hiroshi; Ryu, Kouichi; Yoshimura, Yoshikane

    2001-11-15

    The new burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) is proposed. With this burnup strategy, distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes. The excess reactivity is constant during the burnup. Therefore, any control mechanisms for the burnup are not required. Calculation procedures are presented to find these shapes and the speed of the burning region with the neutron multiplication factor of a reactor employing this burnup strategy.To demonstrate the CANDLE burnup strategy, it is applied to a fast reactor with excellent neutron economy. Only the initially built reactor requires some fissile material such as plutonium or enriched uranium for the nuclear ignition region of its core, but only natural uranium or depleted uranium is required for the other region. Succeeding reactors require only natural or depleted uranium since the burning region of the previous reactor can be utilized for the ignition region. The life of a reactor can be made longer by elongating the core height. The drift speed of the burning region for the presented fast reactor design is {approx}4 cm/yr, which is a preferable value for designing a long-life reactor. The burnup of spent fuel is {approx}40%. It is equivalent to 40% utilization of natural uranium without reprocessing and enrichment.

  1. Designing Critical Experiments in Support of Full Burnup Credit

    SciTech Connect

    Mueller, Don; Roberts, Jeremy A

    2008-01-01

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

  2. Final LDRD report : nanoscale mechanisms in advanced aging of materials during storage of spent %22high burnup%22 nuclear fuel.

    SciTech Connect

    Clark, Blythe G.; Rajasekhara, Shreyas; Enos, David George; Dingreville, Remi Philippe Michel; Doyle, Barney Lee; Hattar, Khalid Mikhiel; Weiner, Ruth F.

    2013-09-01

    We present the results of a three-year LDRD project focused on understanding microstructural evolution and related property changes in Zr-based nuclear cladding materials towards the development of high fidelity predictive simulations for long term dry storage. Experiments and modeling efforts have focused on the effects of hydride formation and accumulation of irradiation defects. Key results include: determination of the influence of composition and defect structures on hydride formation; measurement of the electrochemical property differences between hydride and parent material for understanding and predicting corrosion resistance; in situ environmental transmission electron microscope observation of hydride formation; development of a predictive simulation for mechanical property changes as a function of irradiation dose; novel test method development for microtensile testing of ionirradiated material to simulate the effect of neutron irradiation on mechanical properties; and successful demonstration of an Idaho National Labs-based sample preparation and shipping method for subsequent Sandia-based analysis of post-reactor cladding.

  3. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

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

    NASA Astrophysics Data System (ADS)

    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

  5. Inherent safety of minimum-burnup breed and burn reactors

    SciTech Connect

    Qvist, S.; Reenspan, E.

    2012-07-01

    Reactors that aim to sustain the breed and burn (B and B) mode of operation at minimum discharge burnup require excellent neutron economy, Minimum-burnup B and B cores are generally large and feature low neutron leakage probability and a hard neutron spectrum. While highly promising fuel cycles can be achieved with such designs, the very same features are pushing the limits of the core's ability to passively respond safely to unprotected accidents. Low leakage minimum-burnup sodium-cooled B and B cores have a large positive coolant void-worth and coolant temperature reactivity coefficient. In this study, the applicability of major approaches for fast reactor void-worth reduction is evaluated specifically for B and B cores. The design, shuffling scheme and performance of a new metallic-fueled, sodium-cooled minimum burnup B and B core, used as basis for the void-worth reduction analysis, is presented. The analysis shows that reactivity control systems based on passive {sup 6}Li injection during temperature excursions are the only option able to provide negative void-worth without significantly increasing the minimum burnup required for sustaining the B and B mode of operation. A new type of lithium expansion module (LEM) system was developed specifically for B and B cores and its effect on core performance is presented. (authors)

  6. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    SciTech Connect

    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

  7. Phenomena and Parameters Important to Burnup Credit

    SciTech Connect

    Parks, C.V.

    2001-01-10

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given.

  8. Fuel-Cycle of 'CANDLE' Burnup with Depleted Uranium

    SciTech Connect

    Hiroshi, Sekimoto

    2006-07-01

    A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burnup strategy can derive many merits, especially from safety point of view. The change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40 % of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50 X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The equilibrium core contains a lot of instable materials such as higher actinides and fission products, the enough amounts of which can not be obtained easily. The construction of the initial core is a difficult problem. However, by using enriched uranium substituted for actinides in the equilibrium core, we can construct the initial core whose power profile is similar to the equilibrium one and will reach the equilibrium state without any big change during transient. At present we do not have any material standing for such a high burnup. However, the CANDLE burnup can be realized by employing

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

    SciTech Connect

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

    1997-12-01

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

  10. Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t

    SciTech Connect

    Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R.

    2006-07-01

    High-resolution gamma spectroscopy has been employed for the measurement of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 134}Cs/{sup 154}Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UO{sub 2} pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd/t have been experimentally characterised. Additionally, pin cell depletion calculations have been performed for each sample with the deterministic code CASMO-4, using both its JEF2.2- and its ENDF/B-IV-based libraries, for three different descriptions of the fuel rod irradiation histories, in order to test the sensitivity of the results to neutron cross sections and to the depletion model employed. Measured and calculated ratios have then been compared. It is shown that the {sup 134}Cs/{sup 137}Cs ratio, frequently used as burnup monitor, is considerably less accurate for values exceeding 50 GWd/t; discrepancies of up to {approx}25% are found between measured and calculated values. The ratios built with the {sup 154}Eu concentration show much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by the use of different basic cross section data. (authors)

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

    NASA Astrophysics Data System (ADS)

    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

  12. Fuel Modelling at Extended Burnup: IAEA Coordinated Research Project FUMEX-II

    SciTech Connect

    Killeen, J.C.; Turnbull, J.A.; Sartori, E.

    2007-07-01

    The International Atomic Energy Agency sponsored a Coordinated Research Project on Fuel Modelling at Extended Burnup (FUMEX-II). Eighteen fuel modelling groups participated with the intention of improving their capabilities to understand and predict the behaviour of water reactor fuel at high burnups. The exercise was carried out in coordination with the OECD/NEA. The participants used a mixture of data derived from actual irradiation histories of high burnup experimental fuel and commercial irradiations where post-irradiation examination measurements are available, combined with idealised power histories intended to represent possible future extended dwell commercial irradiations and test code capabilities at high burnup. All participants have been asked to model nine priority cases out of some 27 cases made available to them for the exercise from the IAEA/OECD International Fuel Performance Experimental Database. Calculations carried out by the participants, particularly for the idealised cases, have shown how varying modelling assumptions affect the high burnup predictions, and have led to an understanding of the requirements of future high burnup experimental data to help discriminate between modelling assumptions. This understanding is important in trying to model transient and fault behaviour at high burnup. It is important to recognise that the code predictions presented here should not be taken to indicate that some codes do not perform well. The codes have been designed for different applications and have differing assumptions and validation ranges; for example codes intended to predict Candu fuel operation with thin wall collapsible cladding do not need the clad creep and gap conductivity modelling found in PWR codes. Therefore, when a case is based on Candu technology or PWR technology, it is to be expected that the codes may not agree. However, it is the very differences in such behaviour that is useful in helping to understand the effects of such

  13. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    SciTech Connect

    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).

  14. Detailed Burnup Calculations for Testing Nuclear Data

    NASA Astrophysics Data System (ADS)

    Leszczynski, F.

    2005-05-01

    A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions for the system at successive burnup time steps are thus based on a calculation route where both geometry and cross sections are accurately represented, without geometry simplifications and with continuous energy data, providing an independent approach for benchmarking other methods and nuclear data of actinides, fission products, and other burnable absorbers. The main advantage of this method over the classical deterministic methods currently used is that the MCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burnup codes adopted until now are the widely used MCNP and ORIGEN codes, but other codes can be used also. For using this method, there is need of a well-known set of nuclear data for isotopes involved in burnup chains, including burnable poisons, fission products, and actinides. For fixing the data to be included in this set, a study of the present status of nuclear data is performed, as part of the development of the MCQ method. This study begins with a review of the available cross-section data of isotopes involved in burnup chains for power and research nuclear reactors. The main data needs for burnup calculations are neutron cross sections, decay constants, branching ratios, fission energy, and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross

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

    SciTech Connect

    Enercon Services, Inc.

    2011-03-14

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

  16. Fuel rod and core materials investigations related to LWR extended burnup operation

    NASA Astrophysics Data System (ADS)

    Kolstad, Erik; Vitanza, Carlo

    1992-06-01

    The paper deals with tests and recent measurements related to extended burnup fuel performance and describes test facilities and results in the areas of waterside cladding corrosion and irradiation-assisted stress corrosion cracking (IASCC). Fuel temperature data suggest a gradual degradation of UO 2 thermal conductivity with exposure in the range 6-8% per 10 MWd/kgUO 2 at temperatures below 700°C. The effect on the fuel microstructure of interlinkage and resintering phenomena is shown by measuring the surface-to-volume ( S/ V) ratio of the fuel. Changes in S/V with burnup are correlated to power rating and fuel operating temperature. No evidence was found of enhanced fission gas release during load-follow operation in the burnup range 25-45 MWd/kgUO 2. The effect of high lithium concentration (high pH) on the corrosion behaviour of pre-irradiated high burnup Zircaloy-4 fuel rods subjected either to nucleate boiling or to one-phase cooling conditions was studied. The oxide thickness growth rates measured at an average burnup up to 40 MWd/kgUO 2 are consistent with literature data and show no evidence of corrosion enhancement due to the high lithium content and little effect of cooling regime. A test facility for exploring the effects of environmental variables on IASCC behaviour of in-core structural materials is described.

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

    NASA Astrophysics Data System (ADS)

    Fensin, Michael Lorne

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

  18. Extended Burnup Demonstration Reactor Fuels Program. Annual progress report, April 1983-March 1984. [BWR

    SciTech Connect

    Exarhos, C.A.

    1985-06-20

    The US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities Nuclear Corporation have participated since 1979 in a cooperative Extended Burnup Demonstration Program. Under the program, standard ENC-fabricated reload fuel in the Big Rock Point and Oyster Creek reactor cores has been irradiated to discharge burnups at or beyond 35,000 MWD/MTU, one to two cycles beyond its originally projected exposure life. The program provides for examination of the fuel at poolside before and after each extended burnup cycle as well as for limited destructive hot cell examination. The 1984 progress report covers work performed under the EBD program between April 1983 and March 1984. Major milestones reached during the period include completion of a hot cell examination on four high burnup rods from Big Rock Point and of a poolside on the Oyster Creek EBD fuel at discharge. The hot cell examination of four rods at burnups to 37.2 GWD/MTU confirmed poolside measurements on the same fuel, showing the urania and gadolinia-bearing fuel rods to be in excellent condition. No major cladding degradation, pellet restructuring, or pellet-clad interaction was found in any of the samples examined. The Oyster Creek fuel, examined at an assembly average exposure of 34.5 GWD/MTU, showed good performance with regard to both diametral creepdown and clad oxide accumulation.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Marshall, William BJ J

    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.

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

    DOE PAGESBeta

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

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma

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

    SciTech Connect

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

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry

  3. Revised Burnup Code System SWAT: Description and Validation Using Postirradiation Examination Data

    SciTech Connect

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    2002-05-15

    The burnup code system Step-Wise Burnup Analysis Code System (SWAT) is revised for use in a burnup credit analysis. An important feature of the revised SWAT is that its functions are achieved by calling validated neutronics codes without any changes to the original codes. This feature is realized with a system function of the operating system, which allows the revised SWAT to be independent of the development status of each code.A package of the revised SWAT contains the latest libraries based on JENDL-3.2 and the second version of the JNDC FP library. These libraries allow us to analyze burnup problems, such as an analysis of postirradiation examination (PIE), using the latest evaluated data of not only cross sections but also fission yield and decay constants.Another function of the revised SWAT is a library generator for the ORIGEN2 code, which is one of the most reliable burnup codes. ORIGEN2 users can obtain almost the same results with the revised SWAT using the library prepared by this function.The validation of the revised SWAT is conducted by calculation of the Organization for Economic Cooperation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark Phase I-B and analyses of PIE data for spent fuel from Takahama Unit 3. The analysis of PIE data shows that the revised SWAT can predict the isotopic composition of main uranium and plutonium with a deviation of 5% from experimental results taken from UO{sub 2} fuels of 17 x 17 fuel assemblies. Many results of fission products including samarium are within a deviation of 10%. This means that the revised SWAT has high reliability to predict the isotopic composition for pressurized water reactor spent fuel.

  4. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    SciTech Connect

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

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

    SciTech Connect

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

    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.

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

    SciTech Connect

    DeHart, M.D.; Parks, C.V.; Brady, M.C.

    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.

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

    SciTech Connect

    BSC

    2004-12-01

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

  8. Burnup calculation methodology in the serpent 2 Monte Carlo code

    SciTech Connect

    Leppaenen, J.; Isotalo, A.

    2012-07-01

    This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)

  9. Using NDA Techniques to Improve Safeguards Metrics on Burnup Quantification and Plutonium Content in LWR SNF

    SciTech Connect

    Saavedra, Steven F; Charlton, William S; Solodov, Alexander A; Ehinger, Michael H

    2010-01-01

    Globally, there exists a long history in reprocessing in evaluation of the shipper/receiver difference (SRD) on spent nuclear fuel (SNF) received and processed. Typically, the declared shipper s values for uranium and plutonium in SNF (based on calculations involving the initial manufacturer s data and reactor operating history) are used as the input quantities to the head-end process of the facility. Problems have been encountered when comparing these values with measured results of the input accountability tank contents. A typical comparison yields a systematic bias indicated as a loss of 5 7 percent of the plutonium (Pu) and approximately 1 percent for the uranium (U). Studies suggest that such deviation can be attributed to the non-linear nature of the axial burnup values of the SNF. Oak Ridge National Laboratory and Texas A&M University are co-investigating the development of a new method, via Nondestructive Assay (NDA) techniques, to improve the accuracy in burnup and Pu content quantification. Two major components have been identified to achieve this objective. The first component calculates a measurement-based burnup profile along the axis of a fuel rod. Gamma-ray data is collected at numerous locations along the axis of the fuel rod using a High Purity Germanium (HPGe) detector designed for a wide range of gamma-ray energies. Using two fission products, 137Cs and 134Cs, the burnup is calculated at each measurement location and a profile created along the axis of the rod based on the individual measurement locations. The second component measures the U/Pu ratio using an HPGe detector configured for relatively low-energy gamma-rays including x-rays. Fluorescence x-rays from U and Pu are measured and compared to the U/Pu ratio determined from a destructive analysis of the sample. This will be used to establish a relationship between the measured and actual values. This relationship will be combined with the burnup analysis results to establish a relationship

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

    SciTech Connect

    Kee, E.

    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.

  11. A feasibility study to determine cooling time and burnup of ATR fuel using a nondestructive technique and three types of gamma-ray detectors

    SciTech Connect

    Navarro, J.; Aryaeinejad, R.; Nigg, D.W.

    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)

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

    SciTech Connect

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

    2010-01-01

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

  13. Grain and burnup dependence of spent fuel oxidation: geological repository impact

    SciTech Connect

    Hanson, B. D.; Kansa, E. J.; Stoot, R.B.

    1998-10-15

    Further refinements to the oxidation model of Stout et al. have been made. The present model incorporates the burnup dependence of the oxidation rate in addition to an allowance for a distribution of grain sizes. The model was tested by comparing the model results with the oxidation histories of spent fuel samples oxidized in Thermogravimetric Analysis (TGA) or Oven Dry-Bath (ODB) experiments. The comparison between the experimental and model results are remarkably close and confirm the assumption that grain-size distributions and activation energies are the important parameters to predicting oxidation behavior. The burnup dependence of the activation energy was shown to have a greater effect than decreasing the effective grain size in suppressing the rate of the reaction U{sub 4}O{sub 9}(rightwards arrow)U{sub 3} O{sub 4}. Model results predict that U{sub 3}O{sub 8} formation of spent fuels exposed to oxygen will be suppressed even for high burnup fuels that have undergone restructuring in the rim region, provided the repository temperature is kept sufficient.

  14. Preliminary Study of Burnup Characteristics for a Simplified Small Pebble Bed Reactor

    SciTech Connect

    Irwanto, Dwi; Kato, Yukikata; Obara, Toru; Yamanaka, Ichiro

    2010-06-22

    Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu a Peu (little by little) fueling scheme. In the Peu a Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. The objectives of the the present study were to find a means of carrying out the exact calculations needed to analyze the Peu a Peu fuel-loading scheme and to optimize the fuel composition, and fuel-loading scheme to achieve better burnup characteristics. The Monte Carlo method is used to perform calculations with high accuracy. Before the calculation of the whole core, the analysis for the infinite geometry was performed. The power generated per mass consumed for each combination of the uranium enrichment and packing fraction was analyzed from the parametric survey. By using the optimal value obtained, a whole-core calculation for the small 20 MWth reactor was performed and the criticality and burnup of this design was analyzed.

  15. Preliminary Study of Burnup Characteristics for a Simplified Small Pebble Bed Reactor

    NASA Astrophysics Data System (ADS)

    Irwanto, Dwi; Kato, Yukikata; Yamanaka, Ichiro; Obara, Toru

    2010-06-01

    Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu à Peu (little by little) fueling scheme. In the Peu à Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. The objectives of the the present study were to find a means of carrying out the exact calculations needed to analyze the Peu à Peu fuel-loading scheme and to optimize the fuel composition, and fuel-loading scheme to achieve better burnup characteristics. The Monte Carlo method is used to perform calculations with high accuracy. Before the calculation of the whole core, the analysis for the infinite geometry was performed. The power generated per mass consumed for each combination of the uranium enrichment and packing fraction was analyzed from the parametric survey. By using the optimal value obtained, a whole-core calculation for the small 20 MWth reactor was performed and the criticality and burnup of this design was analyzed.

  16. Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors

    SciTech Connect

    Oggianu, Stella Maris; No, Hee Cheon; Kazimi, Mujid S.

    2003-09-15

    A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO{sub 2}), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO{sub 2}/UO{sub 2}), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials at high burnups. Also, chemical compatibility issues are to be considered separately.The same methodology can be applied in the future to evaluate the economic potential of other nuclear fuel materials including different cladding designs, dispersions of ceramics into ceramics, dispersions of ceramics into metals, and also for geometries other than the traditional circular fuel pin.

  17. Monte-Carlo Continuous Energy Burnup Code System.

    Energy Science and Technology Software Center (ESTSC)

    2007-08-31

    Version 00 MCB is a Monte Carlo Continuous Energy Burnup Code for a general-purpose use to calculate a nuclide density time evolution with burnup or decay. It includes eigenvalue calculations of critical and subcritical systems as well as neutron transport calculations in fixed source mode or k-code mode to obtain reaction rates and energy deposition that are necessary for burnup calculations. The MCB-1C patch file and data packages as distributed by the NEADB are verymore » well organized and are being made available through RSICC as received. The RSICC package includes the MCB-1C patch and MCB data libraries. Installation of MCB requires MCNP4C source code and utility programs, which are not included in this MCB distribution. They were provided with the now obsolete CCC-700/MCNP-4C package.« less

  18. Issues related to criticality safety analysis for burnup credit applications

    SciTech Connect

    DeHart, M.D.; Parks, C.V.

    1995-12-01

    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. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. This paper discusses the results of studies to determine the effect of two important modeling assumptions on the criticality analysis of pressurized-water reactor (PWR) spent fuel: (1) the effect of assumed burnup history (i.e., specific power during and time-dependent variations in operational power) during depletion calculations, and (2) the effect of axial burnup distributions on the neutron multiplication factor calculated for a three-dimensional (3-D) conceptual cask design.

  19. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

    SciTech Connect

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-07-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

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

    SciTech Connect

    Sternat, M.; Nichols, T.

    2011-06-09

    Reactor burnup or depletion codes are used thoroughly in the fields of nuclear forensics and nuclear safeguards. Two common codes include MONTEBURNS and MCNPX/CINDER. These are Monte-Carlo depletion routines utilizing MCNP for neutron transport calculations and either ORIGEN or CINDER for burnup calculations. Uncertainties exist in the MCNP steps, but this information is not passed to the depletion calculations or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of 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 150 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. The distributions for each code are a statistical benchmark and comparisons made. 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 the results appear to not. Statistical analyses are performed using the {chi}{sup 2} test against a normal distribution for the k-effective results and several isotopes including {sup 134}Cs, {sup 137}Cs, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, and {sup 240}Pu.

  1. Need for higher fuel burnup at the Hatch Plant

    SciTech Connect

    Beckhman, J.T.

    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.

  2. Comparison of PWR - Burnup calculations with SCALE 5.0/TRITON other burnup codes and experimental results

    SciTech Connect

    Oberle, P.; Broeders, C. H. M.; Dagan, R.

    2006-07-01

    The increasing tendency towards fuel lifetime extension in thermal nuclear reactors motivated validation work for available evaluation tools for nuclear fuel burnup calculations. In this study two deterministic codes with different transport solvers and one Monte Carlo method are investigated. The code system KAPROS/KARBUS uses the classical deterministic First Collision Probability method utilizing a cylinderized Wigner-Seitz cell. In the SCALES.0/TRITON/NEWT code the Extended Step Characteristic method is applied. In a first step the two deterministic codes are compared with experimental results from the KWO-Isotope Correlation Experiment up to 30 MWD/kg HM burnup, published in 1981. Two pin cell calculations are analyzed by comparison of calculated and experimental results for important heavy isotope vectors. The results are very satisfactory. Subsequently, further validation at higher burnup (< 80 MWD/kg HM) is provided by comparison of the two deterministic codes and the Monte Carlo based burnup code MONTEBURNS for PWR UO{sub 2} fuel assembly calculations. Possible reasons for differences in the results are analyzed and discussed. Especially the influence of cross section data and processing is presented. (authors)

  3. Low Burnup Inert Matrix Fuels Performance: TRANSURANUS Analysis of the Halden IFA-652 First Irradiation Cycle

    SciTech Connect

    Calabrese, R.; Vettraino, F.; Tverberg, T.

    2006-07-01

    The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd.kgU{sub eq}{sup -1} vs. 45 MWd.kgU{sub eq}{sup -1} (40 MWd.kgUOX{sub eq}{sup -1}) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd.kgU{sub eq}{sup -1}) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed. (authors)

  4. Nuclide Importance and the Steady-State Burnup Equation

    SciTech Connect

    Sekimoto, Hiroshi; Nemoto, Atsushi

    2000-05-15

    Conventional methods for evaluating some characteristic values of nuclides relating to burnup in a given neutron spectrum are reviewed in a mathematically systematic way, and a new method based on the importance theory is proposed. In this method, these characteristic values of a nuclide are equivalent to the importances of the nuclide. By solving the equation adjoint to the steady-state burnup equation with a properly chosen source term, the importances for all nuclides are obtained simultaneously.The fission number importance, net neutron importance, fission neutron importance, and absorbed neutron importance are evaluated and discussed. The net neutron importance is a measure directly estimating neutron economy, and it can be evaluated simply by calculating the fission neutron importance minus the absorbed neutron importance, where only the absorbed neutron importance depends on the fission product. The fission neutron importance and absorbed neutron importance are analyzed separately, and detailed discussions of the fission product effects are given for the absorbed neutron importance.

  5. Accuracy considerations for Chebyshev rational approximation method (CRAM) in Burnup calculations

    SciTech Connect

    Pusa, M.

    2013-07-01

    The burnup equations can in principle be solved by computing the exponential of the burnup matrix. However, due to the difficult numerical characteristics of burnup matrices, the problem is extremely stiff and the matrix exponential solution has previously been considered infeasible for an entire burnup system containing over a thousand nuclides. It was recently discovered by the author that the eigenvalues of burnup matrices are generally located near the negative real axis, which prompted introducing the Chebyshev rational approximation method (CRAM) for solving the burnup equations. CRAM can be characterized as the best rational approximation on the negative real axis and it has been shown to be capable of simultaneously solving an entire burnup system both accurately and efficiently. In this paper, the accuracy of CRAM is further studied in the context of burnup equations. The approximation error is analyzed based on the eigenvalue decomposition of the burnup matrix. It is deduced that the relative accuracy of CRAM may be compromised if a nuclide concentration diminishes significantly during the considered time step. Numerical results are presented for two test cases, the first one representing a small burnup system with 36 nuclides and the second one a full a decay system with 1531 nuclides. (authors)

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

    SciTech Connect

    Scaglione, John M; Wagner, John C

    2010-01-01

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

  7. Extended burnup fuel - a beneficial environmental step for the nuclear fuel cycle

    SciTech Connect

    Tulenko, J.S. )

    1992-01-01

    The beneficial environmental effects generated by extending the burnup of light water reactors (LWR) fuel covers the entire fuel cycle from cradle (mining) to grave (fuel disposal). The purpose of this paper is to outline the beneficial efforts of extending burnup and show how a proposed financial incentive waste charge would encourage the development and utilization of extended burnup. Environmental consciousness demands that economic incentives be implemented to encourage utilities to extend burnup of their LWR fuel. This paper builds on earlier work and focuses and extends their results. The cornerstone of the positive environmental effects of extended burnup results from the increased plutonium energy generation as burnup is increased and residence time is increased.

  8. Burn-up and neutron economy of accelerator-driven reactor

    SciTech Connect

    Takahashi, H.; Yang, W.; An, Y.; Yamazaki, Y.

    1997-07-01

    It is desirable to have only a small reactivity change in the large burn-up of a solid fuel fast reactor, so that the number of replacements or shuffling of the fuel can be reduced, and plant factor accordingly increased. Also, this reduces the number of control rods needed for the change in burn-up reactivity. In subcritical operation, power controlled by beam power is suggested, but this practice is not as economical as the use of control rods and makes more careful operation of the accelerator is required due to changes in the wake field. In subcritical operation, even a slightly subcritical one, the safety problems associated with a hard neutron spectrum can be alleviated. Neutron leakage from a flattened core, which is needed for operation of the critical fast reactor can be lessen by using the non flat core which has good neutron economy. For generating nuclear energy, it is essential to have a high neutron economy, although breeding the fuel is not welcomed in the present political climate, as is needed for transmuting long lived fission products. In contrast to the breeder, the accelerator driven reactor can separate the energy production from fuel production and processing. Thus, it is suited for non-proliferation of nuclear material by prohibiting the processing and production of fuel in the unrestricted area so this can be only done in international controlled areas which are restricted and remote.

  9. Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel

    SciTech Connect

    Wagner, J.C.

    2000-10-12

    Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For criticality safety purposes, current PWR storage and transportation canister designs employ flux traps between assemblies. Credit for fuel burnup will eliminate the need for these flux traps, and thus, significantly increase the PWR assembly capacity (for a fixed canister volume). Increases in assembly capacity of approximately one-third are expected. In contrast, current BWR canister designs do not require flux traps for criticality safety, and thus, are already at their maximum capacity in terms of physical storage. Therefore, benefits associated with burnup credit for BWR storage and transportation casks may be limited to increasing the enrichment capacity and

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

    NASA Astrophysics Data System (ADS)

    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. Gamma-Ray Simulated Spectrum Deconvolution of a LaBr₃ 1-in. x 1-in. Scintillator for Nondestructive ATR Fuel Burnup On-Site Predictions

    DOE PAGESBeta

    Navarro, Jorge; Ring, Terry A.; Nigg, David W.

    2015-03-01

    A deconvolution method for a LaBr₃ 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr₃ simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibrationmore » curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.« less

  12. Gamma-Ray Simulated Spectrum Deconvolution of a LaBr₃ 1-in. x 1-in. Scintillator for Nondestructive ATR Fuel Burnup On-Site Predictions

    SciTech Connect

    Navarro, Jorge; Ring, Terry A.; Nigg, David W.

    2015-03-01

    A deconvolution method for a LaBr₃ 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr₃ simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibration curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.

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

    SciTech Connect

    Navarro, Jorge

    2013-12-01

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

  14. An Automated, Multi-Step Monte Carlo Burnup Code System.

    Energy Science and Technology Software Center (ESTSC)

    2003-07-14

    Version 02 MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2,more » the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included.« less

  15. MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis

    SciTech Connect

    Gray S Chang

    2005-04-01

    The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN2, and data process module calculations are then output successively as the code runs. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by UNIX BASH script MCWO. This paper presents the MCWO-calculated results of the RERTR-1 and -2 , and the Weapons-Grade Mixed Oxiide fuel (Wg-MOX) fuel experiments in ATR and compares the MCWO-calculated results with the measured data.

  16. NFCSim: A Dynamic Fuel Burnup and Fuel Cycle Simulation Tool

    SciTech Connect

    Schneider, Erich A.; Bathke, Charles G.; James, Michael R.

    2005-07-15

    NFCSim is an event-driven, time-dependent simulation code modeling the flow of materials through the nuclear fuel cycle. NFCSim tracks mass flow at the level of discrete reactor fuel charges/discharges and logs the history of nuclear material as it progresses through a detailed series of processes and facilities, generating life-cycle material balances for any number of reactors. NFCSim is an ideal tool for analysis - of the economics, sustainability, or proliferation resistance - of nonequilibrium, interacting, or evolving reactor fleets. The software couples with a criticality and burnup engine, LACE (Los Alamos Criticality Engine). LACE implements a piecewise-linear, reactor-specific reactivity model for its criticality calculations. This model constructs fluence-dependent reactivity traces for any facility; it is designed to address nuclear economies in which either a steady state is never obtained or is a poor approximation. LACE operates in transient and equilibrium fuel management regimes at the refueling batch level, derives reactor- and cycle-dependent initial fuel compositions, and invokes ORIGEN2.x to carry out burnup calculations.

  17. An Automated, Multi-Step Monte Carlo Burnup Code System.

    SciTech Connect

    TRELLUE, HOLLY R.

    2003-07-14

    Version 02 MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2, the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included.

  18. Burnup estimation of fuel sourcing radioactive material based on monitored Cs and Pu isotopic activity ratios in Fukushima N. P. S. accident

    SciTech Connect

    Yamamoto, T.; Suzuki, M.; Ando, Y.

    2012-07-01

    After the severe core damage of Fukushima Dai-Ichi Nuclear Power Station, radioactive material leaked from the reactor buildings. As part of monitoring of radioactivity in the site, measurements of radioactivity in soils at three fixed points have been performed for {sup 134}Cs and {sup 137}Cs with gamma-ray spectrometry and for Pu, Pu, and {sup 240}Pu with {alpha}-ray spectrometry. Correlations of radioactivity ratios of {sup 134}Cs to {sup 137}Cs, and {sup 238}Pu to the sum of {sup 239}Pu and {sup 240}Pu with fuel burnup were studied by using theoretical burnup calculations and measurements on isotopic inventories, and compared with the Cs and Pu radioactivity rations in the soils. The comparison indicated that the burnup of the fuel sourcing the radioactivity was from 18 to 38 GWd/t, which corresponded to that of the fuel in the highest power and, therefore, the highest decay heat in operating high-burnup fueled BWR cores. (authors)

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

    SciTech Connect

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

    2010-01-01

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

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

    SciTech Connect

    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

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

    NASA Astrophysics Data System (ADS)

    Yilmaz, Mine Ozdemir

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

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

    NASA Astrophysics Data System (ADS)

    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

  3. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    SciTech Connect

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-09-30

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.

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

    SciTech Connect

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

    2011-01-01

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

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

    SciTech Connect

    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.

  6. Burnup calculations for KIPT accelerator driven subcritical facility using Monte Carlo computer codes-MCB and MCNPX.

    SciTech Connect

    Gohar, Y.; Zhong, Z.; Talamo, A.; Nuclear Engineering Division

    2009-06-09

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is {approx}375 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the

  7. Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

    SciTech Connect

    Winston, Philip Lon; Sterbentz, James William

    2001-04-01

    Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements’ burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element’s reported burnup or provide a burnup estimate for an element with an unknown burnup.

  8. Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

    SciTech Connect

    Winston, P.L.; Sterbentz, J.W.

    2002-07-01

    Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements' burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element's reported burnup or provide a burnup estimate for an element with an unknown burnup. (authors)

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

    SciTech Connect

    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 loading criteria

  10. Feasibility studies on the burnup measurement of fuel pebbles with HPGe gamma spectrometer

    NASA Astrophysics Data System (ADS)

    Yan, Wei-Hua; Zhang, Li-Guo; Zhang, Zhao; Xiao, Zhi-Gang

    2013-06-01

    The feasibility of utilizing a High Purity Germanium (HPGe) detector for the fuel element burnup measurement in a future Modular Pebble Bed Reactor (MPBR) was studied. First, the HPGe spectrometer was set-up for running the detector at high count rates while keeping the energy resolution adequately high to discriminate the Cs-137 peak from other interfering peaks. Based on these settings, the geometrical conditions are settled. Next, experiments were performed with Co-60 and Cs-137 sources to mimic the counting rates in real applications. With the aid of KORIGEN and MCNP/G4 simulations, it was demonstrated that the uncertainty of the Cs-137 counting rate can be well controlled within 3.5%. Finally, a full size prototype was tested in comparison with detailed Monte Carlo simulation and the efficiency transfer method was further utilized for efficiency calibration. To reduce the uncertainty in the efficiency transfer process, a standard point source embedded in a graphite sphere was used for efficiency calibration. The correction factor due to pebble self-attenuation was carefully studied.

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

    NASA Astrophysics Data System (ADS)

    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

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

    SciTech Connect

    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.

  13. Investigation of the Effect of Fixed Absorbers on the Reactivity of PWR Spent Nuclear Fuel for Burnup Credit

    SciTech Connect

    Wagner, John C.; Sanders, Charlotta E.

    2002-08-15

    The effect of fixed absorbers on the reactivity of pressurized water reactor (PWR) spent nuclear fuel (SNF) in support of burnup-credit criticality safety analyses is examined. A fuel assembly burned in conjunction with fixed absorbers may have a higher reactivity for a given burnup than an assembly that has not used fixed absorbers. As a result, guidance on burnup credit, issued by the U.S. Nuclear Regulatory Commission's Spent Fuel Project Office, recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommendation eliminates a large portion of the currently discharged SNF from loading in burnup credit casks and thus severely limits the practical usefulness of burnup credit. Therefore, data are needed to support the extension of burnup credit to additional SNF. This research investigates the effect of various fixed absorbers, including integral burnable absorbers, burnable poison rods, control rods, and axial power shaping rods, on the reactivity of PWR SNF. Trends in reactivity with relevant parameters (e.g., initial fuel enrichment, burnup and absorber type, exposure, and design) are established, and anticipated reactivity effects are quantified. Where appropriate, recommendations are offered for addressing the reactivity effects of the fixed absorbers in burnup-credit safety analyses.

  14. 60Co as AN On-Line Burnup Indicator for Multi-Pass Pebble Bed Reactors

    NASA Astrophysics Data System (ADS)

    Hawari, Ayman I.; Chen, Jianwei

    2003-06-01

    Multi-pass pebble bed reactor concepts are characterized by circulating fuel systems that cycle the pebbles in and out of the core until the burnup limit is reached. Currently modular designs of such reactors, with nominal powers of approximately 300 MW-thermal, are under consideration for deployment internationally. A concern of the proposed designs is the ability to perform online measurements of the fuel burnup to determine whether a pebble has reached its end-of-life burnup limit (~ 80,000 MWD/MTU). In this work, computational simulations were performed to assess the utilization of a passive gamma ray spectrometric approach to perform this task. However, in addition to using the inherent signatures of the irradiated fuel, the use of the 59Co(n,γ)60Co reaction as a burnup indicator is considered. The results show that the activity ratio of 134Cs/60Co can provide an indicator that is accurate to within 5% at burnup greater than 20,000 MWD/MTU as the power is varied between 50% and 200% of the reactor's thermal power.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

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

    2015-01-01

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

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

    SciTech Connect

    Kucukboyaci, Vefa; Marshall, William BJ J

    2012-01-01

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

  18. Sensitivity and Uncertainty Analysis to Burn-up Estimates on ADS Using ACAB Code

    SciTech Connect

    Cabellos, O; Sanz, J; Rodriguez, A; Gonzalez, E; Embid, M; Alvarez, F; Reyes, S

    2005-02-11

    Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and the dominant chains that contribute most significantly to the uncertainty. The Monte Carlo methodology gives the burnup uncertainty estimates due to the synergetic/global effect of the complete set of XS uncertainties. These uncertainty estimates are valuable to assess the need of any experimental or systematic reevaluation of some uncertainty XSs for ADS.

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

    PubMed

    Wang, T K; Peir, J J

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio. PMID:10670930

  20. Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code

    SciTech Connect

    Cabellos, O.; Sanz, J.; Rodriguez, A.; Gonzalez, E.; Embid, M.; Alvarez, F.; Reyes, S.

    2005-05-24

    Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and the dominant chains that contribute most significantly to the uncertainty. The Monte Carlo methodology gives the burnup uncertainty estimates due to the synergetic/global effect of the complete set of XS uncertainties. These uncertainty estimates are valuable to assess the need of any experimental or systematic re-evaluation of some uncertainty XSs for ADS.

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

    NASA Astrophysics Data System (ADS)

    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

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

    SciTech Connect

    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.

  3. Characterization and irradiation program: extended-burnup gadolina lead test assembly (Mark GdB)

    SciTech Connect

    Newman, L W

    1984-11-01

    This advanced fuel assembly uses a UO/sub 2/-Gd/sub 2/O/sub 3/ fuel burnable-absorber mixture along with other state-of-the-art fuel performance and uranium utilization-enhancing design features that include annular pellets, annealed guide tubes, Zircaloy intermediate grids, and a removable upper end fitting. Goal of the program is to extend the burnup of pressurized water reactor fuel assemblies to 50,000 MWd/mtU batch average burnup. To achieve this goal, five lead test assemblies have been designed, manufactured, characterized, and inserted for irradiation in Oconee Unit 1 cycle 8. One lead test assembly will receive 13,989 MWd/mtU burnup by its discharge at the end of cycle 8. Three of the four remaining lead test assemblies will receive approximately 45,000 MWd/mtU burnup by their discharge at the end of cycle 10. The fourth lead test assembly will receive approximately 58,000 MWd/mtU before being discharged at the end of cycle 11. The lead test assemblies and their constituent components have been extensively characterized to acquire a beginning-of-life data base to compare with future post-irradiation examination results and thereby determine the irradiation performance of the assemblies and their components. This report contains a description of the lead test assemblies and the pre-irradiation characterization data. Also, the plans for irradiating and examining these assemblies are discussed.

  4. An efficient implementation of the Chebyshev Rational Approximation Method (CRAM) for solving the burnup equations

    SciTech Connect

    Pusa, M.; Leppaenen, J.

    2012-07-01

    The Chebyshev Rational Approximation Method (CRAM) has been recently introduced by the authors for solving the burnup equations with excellent results. This method has been shown to be capable of simultaneously solving an entire burnup system with thousands of nuclides both accurately and efficiently. The method was prompted by an analysis of the spectral properties of burnup matrices and it can be characterized as the best rational approximation on the negative real axis. The coefficients of the rational approximation are fixed and have been reported for various approximation orders. In addition to these coefficients, implementing the method only requires a linear solver. This paper describes an efficient method for solving the linear systems associated with the CRAM approximation. The introduced direct method is based on sparse Gaussian elimination where the sparsity pattern of the resulting upper triangular matrix is determined before the numerical elimination phase. The stability of the proposed Gaussian elimination method is discussed based on considering the numerical properties of burnup matrices. Suitable algorithms are presented for computing the symbolic factorization and numerical elimination in order to facilitate the implementation of CRAM and its adoption into routine use. The accuracy and efficiency of the described technique are demonstrated by computing the CRAM approximations for a large test case with over 1600 nuclides. (authors)

  5. BOXER: Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations

    Energy Science and Technology Software Center (ESTSC)

    2010-02-01

    Neutron transport, calculation of multiplication factor and neutron fluxes in 2-D configurations: cell calculations, 2-D diffusion and transport, and burnup. Preparation of a cross section library for the code BOXER from a basic library in ENDF/B format (ETOBOX).

  6. Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

    SciTech Connect

    Gauld, I.C.

    2005-08-12

    U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.

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

    SciTech Connect

    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. The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

    SciTech Connect

    Logar, Marjan; Jeraj, Robert; Glumac, Bogdan

    2003-02-15

    It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool k{sub eff} is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, and burnup up to 30%, the pitch has the greatest influence on the multiplication factor k{sub eff}. In the interval in which the pitch was changed, k{sub eff} decreased for up to {approx}0.4 for standard and {approx}0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, k{sub eff} values are smaller for {approx}0.2 ({approx}0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is {approx}0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the k{sub eff} of spent-fuel pool and can be neglected in spent-fuel pool design.

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

    SciTech Connect

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

    SciTech Connect

    Lancaster, D.; Machiels, A.

    2012-07-01

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

  11. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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

  14. Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

    DOE PAGESBeta

    Fensin, Michael Lorne; Umbel, Marissa

    2015-09-18

    Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fissionmore » yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.« less

  15. Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

    SciTech Connect

    Fensin, Michael Lorne; Umbel, Marissa

    2015-09-18

    Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fission yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.

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

    SciTech Connect

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

    2015-01-01

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

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

    SciTech Connect

    Karpushkin, T. Yu.

    2012-12-15

    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.

  18. Burnup of rhodium SPND in VVER-1000: Method for determination of linear energy release by SPND readings

    SciTech Connect

    Kurchenkov, A. Yu.

    2011-12-15

    A method for determination of linear energy release of a VVER fuel assembly near a rhodium self-powered neutron detector (SPND) is described. The dependence of SPND burnup on the charge passing through it is specified.

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

    NASA Astrophysics Data System (ADS)

    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.

  20. THE EFFECT OF BURNUP AND SEPARATION EFFICIENCY ON URANIUM UTILIZATION AND RADIOTOXICITY

    SciTech Connect

    Samuel Bays; Steven Piet

    2001-11-01

    This paper addresses two fundamental issues of fuel cycle sustainability. The two primary issues of interest are efficient use of the natural uranium resource (cradle), and management of nuclear waste radiotoxicity (grave). Both uranium utilization and radiotoxicity are directly influenced by the burnup achieved during irradiation (transmutation related) and where applicable the separation efficiency (partitioning related). Burnup influences the in-growth of transuranics by breeding them into the fuel cycle. Transuranic breeding is virtually essential to resource sustainability because it increases utilization of naturally abundant fertile U-238. However, the direct consequence of this build-up is the in-growth of transuranic isotopes which generally increase the source of future geologically committed radiotoxicity. For scenarios involving recycle, separation efficiency influences the degree to which this transuranic source term is removed from active service in the fuel stream and made a disposal legacy of human activity.

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

    SciTech Connect

    Nazirzadeh, M.; Ghasemizad, A.; Khanbabei, B.

    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.

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

    NASA Astrophysics Data System (ADS)

    Nazirzadeh, M.; Ghasemizad, A.; Khanbabei, B.

    2015-12-01

    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.

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

    SciTech Connect

    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.

  4. Thermal properties of U-Mo alloys irradiated to moderate burnup and power

    NASA Astrophysics Data System (ADS)

    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 3.30 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.52 × 1021 fissions cm-3 from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  5. Evaluation of burnup credit for fuel storage analysis -- Experience in Spain

    SciTech Connect

    Conde, J.M.; Recio, M.

    1995-04-01

    Several Spanish light water reactor commercial nuclear power plants are close to maximum spent-fuel pool storage capacity. The utilities are working on the implementation of state-of-the-art methods to increase the storage capacity, including both changes in the pool design (recracking) and the implementation of new analysis approaches with reduced conservation (burnup credit). Burnup credit criticality safety analyses have been approved for two pressurized water reactor plants (four units) and one boiling water reactor (BWR); an other BWR storage analysis is being developed at this moment. The elimination of the ``fresh fuel assumption`` increases the complexity of the criticality analysis to be performed, sometimes putting into question the capability of the analytic tools to properly describe this new situation and increasing the scope of the scenarios to be analyzed. From a regulatory perspective, the reactivity reduction associated with burnup of the fuel can be given credit only if the exposure of each fuel bundle can be known with enough accuracy. Subcriticality of spent-fuel storage depends mainly on the initial fuel enrichment, storage geometry, fuel exposure history, and cooling time. The last two aspects introduced new uncertainties in the criticality analysis that should be quantified in an adequate way. In addition, each and every fuel bundle has its own specific exposure history, so that strong assumptions and simplified calculational schemes have to be developed to undertake the analysis. The Consejo de Seguridad Nuclear (CSN), Spanish regulatory authority on the matter of nuclear safety and radiation protection, plays an active role in the development of analysis methods to support burnup credit, making proposals that may be beneficial in terms of risk and cost while keeping the widest safety margins possible.

  6. A Genesis breakup and burnup analysis in off-nominal Earth return and atmospheric entry

    NASA Technical Reports Server (NTRS)

    Salama, Ahmed; Ling, Lisa; McRonald, Angus

    2005-01-01

    The Genesis project conducted a detailed breakup/burnup analysis before the Earth return to determine if any spacecraft component could survive and reach the ground intact in case of an off-nominal entry. In addition, an independent JPL team was chartered with the responsibility of analyzing several definitive breakup scenarios to verify the official project analysis. This paper presents the analysis and results of this independent team.

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

    SciTech Connect

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

    2010-12-23

    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 (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.

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

    NASA Astrophysics Data System (ADS)

    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.

  9. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

    DOE PAGESBeta

    Fensin, M. L.; Galloway, J. D.; James, M. R.

    2015-04-11

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and newmore » predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less

  10. Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

    SciTech Connect

    Parks, C V; DeHart, M D; Wagner, John C

    2000-03-13

    This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan.

  11. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

    SciTech Connect

    Fensin, M. L.; Galloway, J. D.; James, M. R.

    2015-04-11

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and new predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.

  12. Uncertainties in the effects of burnup and their impact on criticality safety licensing criteria

    SciTech Connect

    Carlson, R.W.; Fisher, L.E.

    1990-07-13

    Current criteria for criticality safety for spent fuel shipping and storage casks are conservative because no credit is permitted for the effects of burnup of the fuel inside the cask. Cask designs that will transport and store large numbers of fuel assemblies (20 or more) must devote a substantial part of their payload to criticality control measures if they are to meet this criteria. The Department of Energy is developing the data necessary to support safety analyses that incorporate the effects of burnup for the next generation of spent fuel shipping casks. The efforts described here are devoted to the development of acceptance criteria that will be the basis for accepting safety analyses. Preliminary estimates of the uncertainties of the effects of burnup have been developed to provide a basis for the consideration of critically safety criteria. The criticality safety margins in a spent fuel shipping or storage cask are dominated by the portions of a fuel assembly that are in low power regions of a reactor core, and the reactor operating conditions are very different from spent fuel storage or transport cask conditions. Consequently, the experience that has been gathered during years of reactor operation does not apply directly to the prediction of criticality safety margins for spent fuel shipping or storage casks. The preliminary estimates of the uncertainties presented in this paper must be refined by both analytical and empirical studies that address both the magnitude of the uncertainties and their interdependence. 9 refs., 5 figs.

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

    SciTech Connect

    Wagner, J. C.

    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.

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

    SciTech Connect

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

    SciTech Connect

    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.

  16. Determining Spent Nuclear Fuel's Plutonium Content, Initial Enrichment, Burnup, and Cooling Time

    SciTech Connect

    Cheatham, Jesse R; Francis, Matthew W

    2011-01-01

    The Next Generation of Safeguards Initiative is examining nondestructive assay techniques to determine the total plutonium content in spent nuclear fuel. The goal of this research was to develop new techniques that can independently verify the plutonium content in a spent fuel assembly without relying on an operator's declarations. Fundamentally this analysis sought to answer the following questions: (1) do spent fuel assemblies contain unique, identifiable isotopic characteristics as a function of their burnup, cooling time, and initial enrichment; (2) how much variation can be seen in spent fuel isotopics from similar and dissimilar reactor power operations; and (3) what isotopes (if any) could be used to determine burnup, cooling time, and initial enrichment? To answer these questions, 96,000 ORIGEN cases were run that simulated typical two-cycle operations with burnups ranging from 21,900 to 72,000 MWd/MTU, cooling times from 5 to 25 years, and initial enrichments between 3.5 and 5.0 weight percent. A relative error coefficient was determined to show how numerically close a reference solution has to be to another solution for the two results to be indistinguishable. By looking at the indistinguishable solutions, it can be shown how a precise measurement of spent fuel isotopics can be inconclusive when used in the absence of an operator's declarations. Using this Method of Indistinguishable Solutions (MIS), we evaluated a prominent method of nondestructive analysis - gamma spectroscopy. From this analysis, a new approach is proposed that demonstrates great independent forensic examination potential for spent nuclear fuel by examining both the neutron emissions of Cm-244 and the gamma emissions of Cs-134 and Eu-154.

  17. Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

    SciTech Connect

    G. S. Chang

    2005-08-01

    A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data.

  18. Antineutrinos for Reactor Safeguards: Effect of Fuel Loading and Burnup on the Signal

    NASA Astrophysics Data System (ADS)

    Erickson, Anna; Bernstein, Adam; Bowden, Nathaniel

    2014-02-01

    Various types of nuclear reactor related information, including relative power level and fuel evolution parameters, can be inferred remotely using antineutrino detectors. We show that it is possible to verify assembly-level burnup using information derived from an antineutrino detector if the nominal reactor fuel loading is known. Alternatively, if the core power is measured using an independent method, for example, a thermal hydraulic element, and the nominal core behavior is known, the antineutrino detector has a capability to determine previously unknown MOX loading in the core.

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

    SciTech Connect

    Ranta-aho, A.

    2006-07-01

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

  20. Advanced Fuel Cycle Initiative - Projected Linear Heat Generation Rate and Burnup Calculations

    SciTech Connect

    Richard G. Ambrosek; Gray S. Chang; Debbie J. Utterbeck

    2005-02-01

    This report provides documentation of the physics analysis performed to determine the linear heat generation rate (LHGR) and burnup calculations for the Advanced Fuel Cycle Initiative (AFCI) tests, AFC-1D, AFC-1H, and AFC-1G. The AFC-1D and AFC-1H tests consists of low-fertile metallic fuel compositions and the AFC-1G test consists of non-fertile and low-fertile nitride compositions. These tests will be irradiated in the East Flux Trap (EFT) positions E1, E2, and E3, respectively, during Advanced Test Reactor (ATR) Cycle 135B.

  1. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  2. A validated methodology for evaluating burnup credit in spent fuel casks

    SciTech Connect

    Brady, M.C. ); Sanders, T.L. )

    1991-01-01

    The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in k{sub eff}. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. 24 refs., 3 tabs.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    Li, Cheng-Yue; Deng, Bai-Quan; Huang, Jin-Hua; Yan, Jian-Cheng

    2006-08-01

    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 317 g 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 1.18 kg that is more than the least required amount of tritium storage to start up three of FEB-like fusion reactors.

  5. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

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

    SciTech Connect

    Widiawati, Nina Su’ud, Zaki

    2015-09-30

    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.

  7. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    DOE PAGESBeta

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application ofmore » the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

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

    SciTech Connect

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

    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

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

    NASA Astrophysics Data System (ADS)

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

    2014-12-01

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

  10. Fission gas release from oxide fuels at high burnups (AWBA development program)

    SciTech Connect

    Dollins, C.C.

    1981-02-01

    The steady state gas release, swelling and densification model previously developed for oxide fuels has been modified to accommodate the slow transients in temperature, temperature gradient, fission rate and pressure that are encountered in normal reactor operation. The gas release predictions made by the model were then compared to gas release data on LMFBR-EBRII fuels obtained by Dutt and Baker and reported by Meyer, Beyer, and Voglewede. Good agreement between the model and the data was found. A comparison between the model and three other sets of gas release data is also shown, again with good agreement.

  11. HTGR Reactor Physics and Burnup Calculations Using the Serpent Monte Carlo Code

    SciTech Connect

    Leppanen, Jaakko; DeHart, Mark D

    2009-01-01

    One of the main advantages of the continuous-energy Monte Carlo method is its versatility and the capability to model any fuel or reactor configuration without major approximations. This capability becomes particularly valuable in studies involving innovative reactor designs and next-generation systems, which often lie beyond the capabilities of deterministic LWR transport codes. In this study, a conceptual prismatic HTGR fuel assembly was modeled using the Serpent Monte Carlo reactor physics burnup calculation code, under development at VTT Technical Research Centre of Finland since 2004. A new explicit particle fuel model was developed to account for the heterogeneity effects. The results are compared to other Monte Carlo and deterministic transport codes and the study also serves as a test case for the modules and methods in SCALE 6.

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

    SciTech Connect

    Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

    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.

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

    NASA Astrophysics Data System (ADS)

    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

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

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

    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

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

    SciTech Connect

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

    2011-01-01

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

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

    SciTech Connect

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

    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

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

    SciTech Connect

    Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Martinez, J. S.

    2015-01-01

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

  19. Use of Burnup Credit as a Safety Factor in Handling of NIST Fuel Assemblies in the L Basin of SRS

    SciTech Connect

    Eghbali, DA

    2004-01-07

    Burnup credit was recently used for the first time in criticality safety analysis to support the handling of the National Institute of Standards and Technology spent fuel assemblies in the L Basin of Savannah River Site. Previous criticality safety analyses were based on the fissile content of fresh, unirradiated fuel assemblies, resulting in handling of a group of 10 or less fuel assemblies at a time. Using burnup credit, it was demonstrated that an isolated configuration of up to 14 NITS fuel assemblies, the maximum number of fuel assemblies in a full basket, submerged in a concrete-lined, water-filled pool is subcritical, resulting in several administrative controls being modified or eliminated without compromising safety.

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

    SciTech Connect

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

    2010-06-22

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    Skutnik, Steven E.; 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 134Cs, 137Cs, 154Eu, 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., 134Cs with 137Cs and 154Eu) in conjunction with gross neutron counting (via 244Cm) 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.

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

    NASA Astrophysics Data System (ADS)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

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

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

    SciTech Connect

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-31

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996. 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.

  5. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

  6. Fission-gas release from uranium nitride at high fission rate density

    NASA Technical Reports Server (NTRS)

    Weinstein, M. B.; Kirchgessner, T. A.; Tambling, T. N.

    1973-01-01

    A sweep gas facility has been used to measure the release rates of radioactive fission gases from small UN specimens irradiated to 8-percent burnup at high fission-rate densities. The measured release rates have been correlated with an equation whose terms correspond to direct recoil release, fission-enhanced diffusion, and atomic diffusion (a function of temperature). Release rates were found to increase linearly with burnups between 1.5 and 8 percent. Pore migration was observed after operation at 1550 K to over 6 percent burnup.

  7. In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

    2008-04-16

    A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.

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

    NASA Astrophysics Data System (ADS)

    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.

  9. Applications of high resolution ICP-AES in the nuclear industry

    SciTech Connect

    Johnson, S.G.; Giglio, J.J.; Goodall, P.S.; Cummings, D.G.

    1998-07-01

    Application of high resolution ICP-AES to selected problems of importance in the nuclear industry is a growing field. The advantages in sample preparation time, waste minimization and equipment cost are considerable. Two examples of these advantages are presented in this paper, burnup analysis of spent fuel and analysis of major uranium isotopes. The determination of burnup, an indicator of fuel cycle efficiency, has been accomplished by the determination of {sup 139}La by high resolution inductively coupled plasma atomic emission spectroscopy (HR-ICP-AES). Solutions of digested samples of reactor fuel rods were introduced into a shielded glovebox housing an inductively coupled plasma (ICP) and the resulting atomic emission transmitted to a high resolution spectrometer by a 31 meter fiber optic bundle. Total and isotopic U determination by thermal ionization mass spectrometry (TIMS) is presented to allow for the calculation of burnup for the samples. This method of burnup determination reduces the time, material, sample handling and waste generated associated with typical burnup determinations which require separation of lanthanum from the other fission products with high specific activities. Work concerning an alternative burnup indicator, {sup 236}U, is also presented for comparison. The determination of {sup 235}U:{sup 238}U isotope ratios in U-Zr fuel alloys is also presented to demonstrate the versatility of HR-ICP-AES.

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

    SciTech Connect

    Dieudonne, C.; Dumonteil, E.; Malvagi, F.; Diop, C. M.

    2013-07-01

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

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

    SciTech Connect

    Holly R. Trellue

    1998-12-01

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

  12. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    SciTech Connect

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; Fulwyler, James

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  13. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    DOE PAGESBeta

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; et al

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

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

    DOE PAGESBeta

    Scaglione, John M.; Mueller, Don E.; Wagner, John C.

    2014-12-01

    One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methodsmore » and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth within the

  15. Preliminary TRIGA fuel burn-up evaluation by means of Monte Carlo code and computation based on total energy released during reactor operation

    SciTech Connect

    Borio Di Tigliole, A.; Bruni, J.; Panza, F.; Alloni, D.; Cagnazzo, M.; Magrotti, G.; Manera, S.; Prata, M.; Salvini, A.; Chiesa, D.; Clemenza, M.; Pattavina, L.; Previtali, E.; Sisti, M.; Cammi, A.

    2012-07-01

    Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the Univ. of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation. (authors)

  16. Reactivity and isotopic composition of spent PWR (pressurized-water-reactor) fuel as a function of initial enrichment, burnup, and cooling time

    SciTech Connect

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.

  17. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    NASA Astrophysics Data System (ADS)

    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.

  18. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    SciTech Connect

    Afifah, Maryam Su’ud, Zaki; Miura, Ryosuke; Takaki, Naoyuki; Sekimoto, H.

    2015-09-30

    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.

  19. Stereological evolution of the rim structure in PWR-fuels at prolonged irradiation: Dependencies with burn-up and temperature

    NASA Astrophysics Data System (ADS)

    Spino, J.; Stalios, A. D.; Santa Cruz, H.; Baron, D.

    2006-08-01

    The stereology of the rim-structure was studied for PWR-fuels up to the ninth irradiation cycle, achieving maximum local burn-ups of 240 GWd/tM and beyond. At intermediate radial positions (0.55 < r/ r0 < 0.7), a small increase of the pore and grain size of recrystallized areas was found, which is attributed to the increase of the irradiation temperatures in the outer half-pellet-radius due to deterioration of the thermal conductivity. In the rim-zone marked pore coarsening and pore-density-drop occur on surpassing the local burn-up of 100 GWd/tM, associated with cavity fractions of ≈0.1. Above this threshold the porosity growth rate drops and stabilizes at a value nearing the matrix-gas swelling-rate (≈0.6%/10 GWd/tM). The rim-cavity coarsening shows ingredients of both Ostwald-ripening and coalescence mechanisms. Despite individual pore-contact events, no clusters of interconnected pores were observed up to maximum pore fractions checked (≈0.24). The rim-pore-structure is found to be well represented in its lower bound by the model system of random penetrable spheres, with percolation threshold at ϕc = 0.29. Rim-cavities are expected to remain closed at least up to this limit.

  20. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    SciTech Connect

    Khericha, S.T.

    2002-06-30

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

  1. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    SciTech Connect

    Khericha, Soli T

    2002-06-01

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to ~42 GWd/MT burnup (+ 2.5% as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: ~50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies (@ ~40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches ~40 GWd/MT burnup per MCNP-predicted values.

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

    SciTech Connect

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

    2011-01-01

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

  3. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    SciTech Connect

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

    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.

  4. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    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.

  6. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    DOE PAGESBeta

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Schwalbach, P.; Sjoland, A.; Tobin, Stephen J.; Trellue, Holly; et al

    2016-02-26

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

  7. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    NASA Astrophysics Data System (ADS)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  8. Potential Impact of Interfacial Bonding Efficiency on High-Burnup Spent Nuclear Fuel Vibration Integrity during Normal Transportation

    SciTech Connect

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

    2014-01-01

    Finite element analysis (FEA) was used to investigate the impacts of interfacial bonding efficiency at pellet pellet and pellet clad interfaces on spent nuclear fuel (SNF) vibration integrity. The FEA simulation results were also validated and benchmarked with reverse bending fatigue test results on surrogate rods consisting of stainless steel (SS) tubes with alumina-pellet inserts. Bending moments (M) are applied to the FEA models to evaluate the system responses of the surrogate rods. From the induced curvature, , the flexural rigidity EI can be estimated as EI=M/ . The impacts of interfacial bonding efficiency on SNF vibration integrity include the moment carrying capacity distribution between pellets and clad and the impact of cohesion on the flexural rigidity of the surrogate rod system. The result also indicates that the immediate consequences of interfacial de-bonding are a load carrying capacity shift from the fuel pellets to the clad and a reduction of the composite rod flexural rigidity. Therefore, the flexural rigidity of the surrogate rod and the bending moment bearing capacity between the clad and fuel pellets are strongly dependent on the efficiency of interfacial bonding at the pellet pellet and pellet clad interfaces. The above-noted phenomenon was calibrated and validated by reverse bending fatigue testing using a surrogate rod system.

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

    NASA Astrophysics Data System (ADS)

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

    2015-07-01

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

  10. An intrinsically safe facility for forefront research and training on nuclear technologies — Burnup and transmutation

    NASA Astrophysics Data System (ADS)

    Lomonaco, G.; Frasciello, O.; Osipenko, M.; Ricco, G.; Ripani, M.

    2014-04-01

    The currently dominant open fuel cycles have resulted in the gradual accumulation of (relatively) large quantities of highly radioactive or fertile materials in the form of depleted uranium, plutonium, minor actinides (MA) and long-lived fission products (LLFP). For low-activity wastes a heavily shielded surface repository is required. Spent fuel can be instead directly buried in deep geological repositories or reprocessed in order to separate U and Pu and eventually also MA and LLFP from other materials. These elements can be further burnt by modern reactors but not yet in sufficient quantities to slow down the steady accumulation of these materials in storage. Using ADS, the residual long-lifetime isotopes can be transmuted by nuclear reactions into shorter-lifetime isotopes again storable in surface repositories. However, in order to perform transmutations at a practical level, high-power reactors (and consequently high-power accelerators) are required; particularly, a significant transmutation can be reached not only by increasing the beam current to something of the order of a few tens of mA, but also by increasing the beam energy above 500MeV in order to reach the spallation regime. Such high-power infrastructures require intermediate test facilities with lower power and higher safety level for the investigation of their dynamics and transmutation capabilities: the ADS proposed in this study could accomplish many of these constraints.

  11. Direct Deterministic Method for Neutronics Analysis and Computation of Asymptotic Burnup Distribution in a Recirculating Pebble-Bed Reactor

    SciTech Connect

    Terry, William Knox; Gougar, Hans D; Ougouag, Abderrafi Mohammed-El-Ami

    2002-07-01

    A new deterministic method has been developed for the neutronics analysis of a pebble-bed reactor (PBR). The method accounts for the flow of pebbles explicitly and couples the flow to the neutronics. The method allows modeling of once-through cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times. This new work is distinguished from older methods by the systematically semi-analytical approach it takes. In particular, whereas older methods use the finite-difference approach (or an equivalent one) for the discretization and the solution of the burnup equation, the present work integrates the relevant differential equation analytically in discrete and complementary sub-domains of the reactor. Like some of the finite-difference codes, the new method obtains the asymptotic fuel-loading pattern directly, without modeling any intermediate loading pattern. This is a significant advantage for the design and optimization of the asymptotic fuel-loading pattern. The new method is capable of modeling directly both the once-through-then-out fuel cycle and the pebble recirculating fuel cycle. Although it currently includes a finite-difference neutronics solver, the new method has been implemented into a modular code that incorporates the framework for the future coupling to an efficient solver such as a nodal method and to modern cross section preparation capabilities. In its current state, the deterministic method presented here is capable of quick and efficient design and optimization calculations for the in-core PBR fuel cycle. The method can also be used as a practical "scoping" tool. It could, for example, be applied to determine the potential of the PBR for resisting nuclear-weapons proliferation and to optimize proliferation-resistant features. However, the purpose of this paper is to show that the method itself is viable. Refinements to the code are under way, with the objective of producing a powerful reactor physics

  12. Direct Measurement of Initial Enrichment, Burn-up and Cooling Time of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    SciTech Connect

    Henzl, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-13

    An outline of this presentation of what a Differential Die-Away (DDA) instrument can do are: (1) Principle of operation of DDA instrument; (2) Determination of initial enrichment (IE) ({sigma} < 5%); (3) Determination of burn up (BU) ({sigma} {approx} 6%); (4) Determination of cooling time (CT) ({sigma} {approx} 20-50%); and (5) DDA instrument as a standalone device. DDA response (fresh fuel vs. spent fuel) is: (1) Fresh fuel => DDA response increases (die-away time is longer) with increasing fissile content; and (2) Spent fuel => DDA response decreases (die-away time is shorter) with higher burn-up (i.e. more neutron absorbers present).

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

    SciTech Connect

    Scaglione, John M.; Mueller, Don E.; Wagner, John C.

    2014-12-01

    One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methods and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth

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

    SciTech Connect

    S. T. Khericha; R. C. Pedersen

    2003-09-01

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

  15. Estimation of average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors by using the {sup 134}Cs/{sup 137}Cs ratio method

    SciTech Connect

    Endo, T.; Sato, S.; Yamamoto, A.

    2012-07-01

    Average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors is estimated, using the {sup 134}Cs/{sup 137}Cs ratio method for measured radioactivities of {sup 134}Cs and {sup 137}Cs in contaminated soils within the range of 100 km from the Fukushima Dai-ichi nuclear power plants. As a result, the measured {sup 134}Cs/{sup 137}Cs ratio from the contaminated soil is 0.996{+-}0.07 as of March 11, 2011. Based on the {sup 134}Cs/{sup 137}Cs ratio method, the estimated burnup of damaged fuels is approximately 17.2{+-}1.5 [GWd/tHM]. It is noted that the numerical results of various calculation codes (SRAC2006/PIJ, SCALE6.0/TRITON, and MVP-BURN) are almost the same evaluation values of {sup 134}Cs/ {sup 137}Cs ratio with same evaluated nuclear data library (ENDF-B/VII.0). The void fraction effect in depletion calculation has a major impact on {sup 134}Cs/{sup 137}Cs ratio compared with the differences between JENDL-4.0 and ENDF-B/VII.0. (authors)

  16. Field test and calibration of neutron coincidence counters for high-mass plutonium samples

    SciTech Connect

    Menlove, H.O.; Dickinson, R.J.; Douglas, I.; Orr, C.; Rogers, F.J.G.; Wells, G.; Schenkel, R.; Smith, G.; Fattgh, A.; Ramalho, A.

    1987-02-01

    Five different neutron coincidence systems were evaluated and calibrated for high-mass PuO/sub 2/ samples. The samples were from 2 to 7.2 kg of PuO/sub 2/ in mass, with a large range of burnup. This report compares the equipment and the results, with an evaluation of deadtime and multiplication corrections.

  17. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

    SciTech Connect

    Wagner, J.C.; Parks, C.V.

    2000-09-01

    This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub inf} estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k{sub inf} estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration ({approximately}2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion ({le} 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of

  18. Direct Measurement of Initial Enrichment and Burn-up of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    SciTech Connect

    Henzl, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-16

    A key objective of the Next Generation Safeguards Initiative (NGSI) is to utilize non-destructive assay (NDA) techniques to determine the elemental plutonium (Pu) content in a commercial-grade nuclear spent fuel assembly (SFA). In the third year of the NGSI Spent Fuel NDA project, the research focus is on the integration of a few NDA techniques. One of the reoccurring challenges to the accurate determination of Pu content has been the explicit dependence of the measured signal on the presence of neutron absorbers which build up in the assembly in accordance with its operating and irradiation history. The history of any SFA is often summarized by the parameters of burn-up (BU), initial enrichment (IE) and cooling time (CT). While such parameters can typically be provided by the operator, the ability to directly measure and verify them would significantly enhance the autonomy of the IAEA inspectorate. Within this paper, we demonstrate that an instrument based on a Differential Die-Away technique is in principle capable of direct measurement of IE and, should the CT be known, also the BU.

  19. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    SciTech Connect

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  20. Using Finite Model Analysis and Out of Hot Cell Surrogate Rod Testing to Analyze High Burnup Used Nuclear Fuel Mechanical Properties

    SciTech Connect

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

    2014-07-01

    Based on a series of FEA simulations, the discussions and the conclusions concerning the impact of the interface bonding efficiency to SNF vibration integrity are provided in this report; this includes the moment carrying capacity distribution between pellets and clad, and the impact of cohesion bonding on the flexural rigidity of the surrogate rod system. As progressive de-bonding occurs at the pellet-pellet interfaces and at the pellet-clad interface, the load ratio of the bending moment carrying capacity gradually shifts from the pellets to the clad; the clad starts to carry a significant portion of the bending moment resistance until reaching the full de-bonding state at the pellet-pellet interface regions. This results in localized plastic deformation of the clad at the pellet-pellet-clad interface region; the associated plastic deformations of SS clad leads to a significant degradation in the stiffness of the surrogate rod. For instance, the flexural rigidity was reduced by 39% from the perfect bond state to the de-bonded state at the pellet-pellet interfaces.

  1. Fuel/cladding compatibility in high-burnup U-19Pu-10Zr/HT9-clad fuel at elevated temperatures

    SciTech Connect

    Cohen, A.B.; Tsai, H.; Neimark, L.A.

    1992-11-01

    This paper summarizes the most recent results of a continuing experimental effort to study compatibility issues of irradiated metallic fuel and cladding at elevated temperatures that may be encountered beyond those of nominal steady-state conditions.

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

    SciTech Connect

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

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

  3. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    SciTech Connect

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  4. Study of UO/sub 2/ wafer fuel for very high-power research reactors

    SciTech Connect

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1980-11-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to < 20% in those research and test reactors presently using highly enriched uranium fuel. UO/sub 2/ caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO/sub 2/ wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full-power operation up to a burnup of 1.09 x 10/sup 21/ fis/cm/sup 3/; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, and wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. 12 figures, 7 tables.

  5. Neutronic assessment of stringer fuel assembly design for liquid-salt-cooledvery high temperature reactor (LS-VHTR).

    SciTech Connect

    Szakaly, F. J.; Kim, T. K.; Taiwo, T. A.

    2006-09-15

    Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.

  6. Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport

    SciTech Connect

    Boshoven, J.K.

    1992-01-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks In 1992. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the fresh fuel'' assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions.

  7. Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport

    SciTech Connect

    Boshoven, J.K.

    1992-08-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks In 1992. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the ``fresh fuel`` assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions.

  8. Development of modified MDA (M-MDA), PWR fuel cladding tube for high duty operation in future

    SciTech Connect

    Watanabe, Seiichi; Kido, Toshiya; Arakawa, Yasushi

    2007-07-01

    A new cladding material of M-MDA has been developed in order to prepare for a strong growing demand for advanced fuel which can maintain its integrity even under high duties due to more efficient operation such as higher burnup, higher LHR, and longer operation cycle which will contribute the suppression of environmental burdens like CO{sub 2} emission. The main aim of M-MDA is to have excellent corrosion resistance while the other properties are inherited from MDA, which has been adopted to the step 2 fuel, instead of Zry-4, of Japanese PWR plant whose upper limit of assembly discharged burnup is 55 MWd/kgU. And we could confirm that the main aim of M-MDA was achieved by means of out-of-pile tests. In order to confirm improvement of corrosion resistance of M-MDA in the actual operation, irradiation test of M-MDA in the commercial reactor of Vandellos II is ongoing. The latest results of on-site examination after every end of cycle showed that oxide thickness of M-MDA-SR was much smaller than that of MDA at rod discharged burnup of approximately 60 MWd/kgU. The final irradiation cycle was completed on April 2007 and then we will obtain corrosion data of M-MDA over 70 MWd/kgU. M-MDA is a candidate alloy for advanced fuel under higher duty usage. (authors)

  9. The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

    SciTech Connect

    Sen, R. S.; Cogliati, J. J.; Gougar, H. D.

    2012-07-01

    The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles, especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the

  10. Nonproliferation issue of the pebble bed high-temperature reactor

    SciTech Connect

    Teuchert, E.; Haas, K.A.

    1986-02-01

    The constraints of nonproliferation of weapons-grade fuel are most favorably observed in the medium enriched uranium (MEU) fuel cycle of the pebble bed high-temperature reactor, using 20% enriched uranium as feed and thorium as breed material. The cycle can be designed so that the uranium enrichment never exceeds the limitation defined for nonsensitive fuel. In the spent fuel, the amount of fissile plutonium is one order of magnitude lower than for the light water reactor and it is strongly denatured by the even-numbered plutonium isotopes. In the once-through option applied in the introductory phase of the reactor, the proliferation restraints of the plutonium are furnished by the choice of the carbon/heavy metal ratio higher than 450 and of the burnup of 100 MWd/kg heavy metal. The Pu/sub FISS/Pu/sub TOTAL/ is achieved as low as 37%, and the admixing of 8% of /sup 238/Pu would complicate its handling by the decay heat rating. In the closed MEU cycle, the /sup 238/U is continuously separated from the cycle by the use of two different types of fuel elements: Thorium and 20% enriched uranium are inserted into the feed elements, and the uranium recovered from the reprocessing is loaded into the burnup elements, without thorium. These elements are removed from the cycle without reprocessing. Again the proliferation risk of the fissile plutonium is minimized because of its very low quantity and high denaturization.

  11. Calculated Neutron and Gamma-ray Spectra across the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    James W. Sterbentz

    2008-05-01

    Neutron and gamma-ray flux spectra are calculated using the MCNP5 computer code and a one-sixth core model of a prismatic Very High Temperature Reactor based on the General Atomics Gas Turbine-Modular Helium Reactor. Spectra are calculated in the five inner reflector graphite block rings, three annular active core fuel rings, three outer graphite reflector block rings, and the core barrel. The neutron spectra are block and fuel pin averages and are calculated as a function of temperature and burnup. Also provided are the total, fast, and thermal radial profile fluxes and core barrel dpa rates.

  12. Performance of HT9 clad metallic fuel at high temperature

    SciTech Connect

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-12-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching {approximately}660{degree}C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area.

  13. Performance of HT9 clad metallic fuel at high temperature

    SciTech Connect

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-01-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching [approximately]660[degree]C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area.

  14. Results of irradiation of (U0.55Pu0.45)N and (U0.4Pu0.6)N fuels in BOR-60 up to ˜12 at.% burn-up

    NASA Astrophysics Data System (ADS)

    Rogozkin, B. D.; Stepennova, N. M.; Fedorov, Yu. Ye.; Shishkov, M. G.; Kryukov, F. N.; Kuzmin, S. V.; Nikitin, O. N.; Belyaeva, A. V.; Zabudko, L. M.

    2013-09-01

    In the article presented are the results of post-irradiation tests of helium bonded fuel pins with mixed mononitride fuel (U0.55Pu0.45)N and (U0.4Pu0.6)N having 85% density irradiated in BOR-60 reactor. Achieved maximum burn-up was, respectively, equal to 9.4 and 12.1 at.% with max linear heat rates 41.9 and 54.5 kW/m. Maximum irradiation dose was 43 dpa. No damage of claddings made of ChS-68 steel (20% cold worked) was observed, and ductility margin existed. Maximum depth of cladding corrosion was within 15 μm. Swelling rates of (U0.4Pu0.6)N and (U0.55Pu0.45)N were, respectively, ˜1.1% and ˜0.68% per 1 at.%. Gas release rate did not exceed 19.3% and 19%. Pattern of porosity distribution in the fuel influenced fuel swelling and gas release rates. Plutonium and uranium are uniformly distributed in the fuel, local minimum values of their content being caused by pores and cracks in the pellets. The observable peaks in content distribution are probably connected with the local formation of isolated phases (e.g. Mo, Pd) while the minimum values refer to fuel pores and cracks. Xenon and cesium tend to migrate from the hot sections of fuel, and therefore their min content is observed in the central section of the fuel pellets. Phase composition of the fuel was determined with X-ray diffractometer. The X-ray patterns of metallographic specimens were obtained by the scanning method (the step was 0.02°, the step exposition was equal to 2 s). From the X-ray diffraction analysis data, it follows that the nitrides of both fuel types have the single-phase structure with an FCC lattice (see Table 6).

  15. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    SciTech Connect

    Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M.

    2012-07-01

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  16. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  17. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    SciTech Connect

    Gauntt, Randall O.; Ross, Kyle W.; Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  18. Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

    NASA Astrophysics Data System (ADS)

    Fratoni, Massimiliano

    This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be

  19. PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

    DOEpatents

    Puechl, K.H.

    1963-09-24

    A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

  20. Nuclear design of small-sized high temperature gas-cooled reactor for developing countries

    SciTech Connect

    Goto, M.; Seki, Y.; Inaba, Y.; Ohashi, H.; Sato, H.; Fukaya, Y.; Tachibana, Y.

    2012-07-01

    Japan Atomic Energy Agency (JAEA) has started a conceptual design of a small-sized HTGR with 50 MW thermal power (HTR50S), which is a first-of-a-kind commercial or demonstration plant of a small-sized HTGR to be deployed in developing countries such as Kazakhstan in the 2020's. The nuclear design of the HTR50S is performed by upgrading the proven technology of the High Temperature Engineering Test Reactor (HTTR) to reduce the cost for the construction. In the HTTR design, twelve kinds of fuel enrichment was used to optimize the power distribution, which is required to make the maximum fuel temperature below the thermal limitation during the burn-up period. However, manufacture of many kinds of fuel enrichment causes increase of the construction cost. To solve this problem, the present study challenges the nuclear design by reducing the number of fuel enrichment to as few as possible. The nuclear calculations were performed with SRAC code system whose validity was proven by the HTTR burn-up data. The calculation results suggested that the optimization of the power distribution was reasonably achieved and the maximum fuel temperature was kept below the limitation by using three kinds of fuel enrichment. (authors)

  1. Neutronics and Depletion Methods for Parametric Studies of Fluoride Salt Cooled High Temperature Reactors with Slab Fuel Geometry and Multi-Batch Fuel Management Schemes

    SciTech Connect

    Cisneros, Anselmo T.; Ilas, Dan

    2013-01-01

    The Advanced High-Temperature Reactor (AHTR) is a 3400 MWth fluoride-salt-cooled high-temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the Reactivity-Equivalent Physical Transformation (RPT) methodology usually applied in systems with coated-particle fuel in cylindrical and spherical geometries has been extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle (BOEC) composition and sampling different discharge burnups. This Iterative Equilibrium Depletion Search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux, and composition evolutions) but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the Non-Linear Reactivity Model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms so that single-batch depletion results could be extrapolated to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.

  2. Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

    2015-03-01

    In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

  3. Neutronics assessment of stringer fuel assembly designs for the liquid-salt-cooled very high temperature reactor (LS-VHTR).

    SciTech Connect

    Szakaly, F. J.; Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2007-01-01

    Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.Evaluations of a liquid-salt- (molten-salt-) cooled version of the prismatic-block type VHTR, the LS-VHTR, are ongoing at U.S. national laboratories, universities, and industry. These evaluations have included core and passive safety studies and balance of plant conceptual designs.

  4. Characterization of mechanical properties and microstructure of highly irradiated SS 316

    NASA Astrophysics Data System (ADS)

    Karthik, V.; Kumar, RanVijay; Vijayaragavan, A.; Venkiteswaran, C. N.; Anandaraj, V.; Parameswaran, P.; Saroja, S.; Muralidharan, N. G.; Joseph, Jojo; Kasiviswanathan, K. V.; Jayakumar, T.; Raj, Baldev

    2013-08-01

    Cold worked austenitic stainless steel type AISI 316 is used as the material for fuel cladding and wrapper of the Fast Breeder Test Reactor (FBTR), India. The evaluation of mechanical properties of these core structurals is very essential to assess its integrity and ensure safe and productive operation of FBTR to very high burn-ups. The changes in the mechanical properties of these core structurals are associated with microstructural changes caused by high fluence neutron irradiation and temperatures of 673-823 K. Remote tensile testing has been used for evaluating the tensile properties of irradiated clad tubes and shear punch test using small disk specimens for evaluating the properties of irradiated hexagonal wrapper. This paper will highlight the methods employed for evaluating the mechanical properties of the irradiated cladding and wrapper and discuss the trends in properties as a function of dpa (displacement per atom) and irradiation temperature.

  5. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  6. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This report presents FY13 activities for the analysis of D2O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D2O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D2O coolant, MOX with different 235U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233U over other fissile like 239Pu and 235U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void reactivity until the target burnup of 51 GWd/t. Several other distinctions between U-Pu and

  7. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    PubMed

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  8. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    NASA Astrophysics Data System (ADS)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-10-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  9. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    PubMed Central

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  10. Fission product release and microstructure changes of irradiated MOX fuel at high temperatures

    NASA Astrophysics Data System (ADS)

    Colle, J.-Y.; Hiernaut, J.-P.; Wiss, T.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    burnups correspond reasonably well with measurement of Walker et al. [11]. All those data are shown Fig. 2.Fragments of 2-8 mg were chosen for the experiments. Since these specimens are small compared to the drilled sample size and were taken randomly, the precise radial position could not be determined, in particular the specimens of sample type, A and B could be from close radial locations.Specimens from each drilled sample type were annealed up to complete vaporisation (˜2600 K) at a speed of about 10 K min-1 in a Knudsen effusion mass spectrometer (KEMS) described previously [13,14]. In addition to helium and to the FGs all the species present in the vapour between 83 and 300 a.m.u. were measured during the heating. Additionally, the 85Kr isotope was analysed in a cold trap by β and γ counting. The long-lived fission gas isotopes correspond to masses 131, 132, 134 and 136 for Xe and 83, 84, 85 and 86 for Kr. The absolute quantities of gas released from specimens of sample types A and B were also determined using the in-house built Q-GAMES (Quantitative gas measurement system), described in detail in [15].For each of the samples, fragments were also annealed and measured in the KEMS up to specific temperatures corresponding to different stages of the FGs or He release. These fragments were subsequently analysed by Scanning Electron Microscopy (SEM, Philips XL40) [16] in order to investigate the relationship between structural changes, burn-up, irradiation temperature and fission products release. SEM observations were also done on the samples before the KEMS experiments and the fracture surface appearance of the samples is shown in Fig. 3, revealing the presence of the high burnup structure (HBS) in the Pu-rich agglomerates.A summary of the 12 samples analysed by KEMS, SEM and Q-GAMES is given in Table 1. At 1300 K no clear change potentially related to gas release appears in the UM and PA. At 1450 K a beginning of grain boundaries opening can be observed as well as

  11. Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

    SciTech Connect

    Michael A. Pope

    2012-07-01

    High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

  12. Space augmentation of military high-level waste disposal

    NASA Technical Reports Server (NTRS)

    English, T.; Lees, L.; Divita, E.

    1979-01-01

    Space disposal of selected components of military high-level waste (HLW) is considered. This disposal option offers the promise of eliminating the long-lived radionuclides in military HLW from the earth. A space mission which meets the dual requirements of long-term orbital stability and a maximum of one space shuttle launch per week over a period of 20-40 years, is a heliocentric orbit about halfway between the orbits of earth and Venus. Space disposal of high-level radioactive waste is characterized by long-term predictability and short-term uncertainties which must be reduced to acceptably low levels. For example, failure of either the Orbit Transfer Vehicle after leaving low earth orbit, or the storable propellant stage failure at perihelion would leave the nuclear waste package in an unplanned and potentially unstable orbit. Since potential earth reencounter and subsequent burn-up in the earth's atmosphere is unacceptable, a deep space rendezvous, docking, and retrieval capability must be developed.

  13. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  14. High Tech High.

    ERIC Educational Resources Information Center

    Hardy, Lawrence

    2001-01-01

    In an era of high-stakes testing and prescriptive teaching styles, a San Diego charter high school embraces project learning, multilevel classrooms, and video portfolios of student work. The school lacks dining, music, and athletic facilities, but features hefty teacher salaries, student freedom, and real-world problem solving. (MLH)

  15. Irradiation testing of high density uranium alloy dispersion fuels

    SciTech Connect

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.

  16. Reactor Physics Studies of Reduced-Tantaulum-Content Control and Safety Elements for the High Flux Isotope Reactor

    SciTech Connect

    Primm, R.T., III

    2003-11-01

    Some of the unirradiated High Flux Isotope Reactor (HFIR) control elements discharged during the late 1990s were observed to have cladding damage--local swelling or blistering. The cladding damage was limited to the tantalum/europium interface of the element and is thought to result from interaction of hydrogen and europium to form a compound of lower density than europium oxide, thus leading to a ''blistering'' of the control plate cladding. Reducing the tantalum loading in the control plates should help preclude this phenomena. The impact of the change to the control plates on the operation of the reactor was assessed. Regarding nominal, steady-state reactor operation, the impact of the change in the power distribution in the core due to reduced tantalum content was calculated and found to be insignificant. The magnitude and impact of the change in differential control element worth was calculated, and the differential worths of reduced tantalum elements vs the current elements from equivalent-burnup critical configurations were determined to be unchanged within the accuracy of the computational method and relevant experimental measurements. The location of the critical control elements symmetric positions for reduced tantalum elements was found to be 1/3 in. less withdrawn relative to existing control elements regardless of the value of fuel cycle burnup (time in the fuel cycle). The magnitude and impact of the change in the shutdown margin (integral rod worth) was assessed and found to be unchanged. Differential safety element worth values for the reduced-tantalum-content elements were calculated for postulated accident conditions and were found to be greater than values currently assumed in HFIR safety analyses.

  17. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  18. Ultrasonic High-Temperature Sensors: Past Experiments and Prospects for Future Use

    NASA Astrophysics Data System (ADS)

    Laurie, M.; Magallon, D.; Rempe, J.; Wilkins, C.; Pierre, J.; Marquié, C.; Eymery, S.; Morice, R.

    2010-09-01

    Ultrasonic thermometry sensors (UTS) have been intensively studied in the past to measure temperatures from 2080 K to 3380 K. This sensor, which uses the temperature dependence of the acoustic velocity in materials, was developed for experiments in extreme environments. Its major advantages, which are (a) capability of measuring a temperature profile from multiple sensors on a single probe and (b) measurement near the sensor material melting point, can be of great interest when dealing with on-line monitoring of high-temperature safety tests. Ultrasonic techniques were successfully applied in several severe accident related experiments. With new developments of alternative materials, this instrument may be used in a wide range of experimental areas where robustness and compactness are required. Long-term irradiation experiments of nuclear fuel to extremely high burn-ups could benefit from this previous experience. After an overview of UTS technology, this article summarizes experimental work performed to improve the reliability of these sensors. The various designs, advantages, and drawbacks are outlined and future prospects for long-term high-temperature irradiation experiments are discussed.

  19. Optimization of Depletion Modeling and Simulation for the High Flux Isotope Reactor

    SciTech Connect

    Betzler, Benjamin R; Ade, Brian J; Chandler, David; Ilas, Germina; Sunny, Eva E

    2015-01-01

    Monte Carlo based depletion tools used for the high-fidelity modeling and simulation of the High Flux Isotope Reactor (HFIR) come at a great computational cost; finding sufficient approximations is necessary to make the use of these tools feasible. The optimization of the neutronics and depletion model for the HFIR is based on two factors: (i) the explicit representation of the involute fuel plates with sets of polyhedra and (ii) the treatment of depletion mixtures and control element position during depletion calculations. A very fine representation (i.e., more polyhedra in the involute plate approximation) does not significantly improve simulation accuracy. The recommended representation closely represents the physical plates and ensures sufficient fidelity in regions with high flux gradients. Including the fissile targets in the central flux trap of the reactor as depletion mixtures has the greatest effect on the calculated cycle length, while localized effects (e.g., the burnup of specific isotopes or the power distribution evolution over the cycle) are more noticeable consequences of including a critical control element search or depleting burnable absorbers outside the fuel region.

  20. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    SciTech Connect

    Paul A. Demkowicz; David V. Laug; Dawn M. Scates; Edward L. Reber; Lyle G. Roybal; John B. Walter; Jason M. Harp; Robert N. Morris

    2012-10-01

    The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 degrees C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated fission gas monitoring system, as well as preliminary system calibration results.

  1. Rapid Transmutation of High-Level Nuclear Wastes in a Catalyzed Fusion-Driven System

    NASA Astrophysics Data System (ADS)

    Demir, Nesrin; Genç, Gamze; Altunok, Taner; Yapıcı, Hüseyin

    2009-03-01

    The aim of this study is to investigate the high-level waste (HLW) transmutation potential of fusion-driven transmuter (FDT) based on catalyzed D-D fusion plasma for various fuel fractions. The Minor actinide (MA) (237Np, 241Am, 243Am and 244Cm) and long-lived fission product (LLFP) (99Tc, 129I and 135Cs) nuclides discharged from high burn-up pressured water reactor-mixed oxide spent fuel are considered as the HLW. The volume fractions of the MA and LLFP are raised from 10 to 20% stepped by 2% and 10 to 80% stepped by 5%, respectively. The transmutation analyses have been performed for an operation period (OP) of up to 6 years by 75% plant factor ( η) under a first-wall neutron load ( P) of 5 MW/m2 by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. The numerical results bring out that the considered FDT has a high neutronic performance for an effective and rapid transmutation of MA and LLFP as well as the energy generation along the OP.

  2. Ultrasonic High-Temperature Sensors: Past Experiments and Prospects for Future Use

    SciTech Connect

    M. Laurie; D. Magallon; J.Rempe; C.Wilkins; C. Marquié; S. Eymery; R. Morice

    2010-08-01

    Ultrasonic thermometry sensors (UTS) have been intensively studied in the past to measure temperatures from 2080 to 3380 K. This sensor, which uses the temperature dependence of acoustic velocity in materials, was developed for experiments in extreme environments. Its major advantages, which are (a) recording capability of a temperature profile deduced from the notches on the sensor rod and (b) measurement near the sensor material melting point, can be of great interest when dealing with on-line monitoring of high temperature safety tests. Ultrasonic techniques were successfully applied in several severe accident related experiments. If new developments are conducted with other materials, this sensor type may be used in a wide-range of experimental areas where robustness and compactness are required. Long-term irradiation experiments of nuclear fuel to extremely high burn-ups could benefit from this previous experience. After an overview of UTS technology, this paper summarizes experimental work performed to improve the reliability of these sensors. The various designs, advantages and drawbacks are outlined and future prospects for long term high temperature irradiation experiments are discussed.

  3. Fuel element concept for long life high power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  4. High-density reduced-enrichment fuels for Research and Test Reactors

    SciTech Connect

    Snelgrove, J.L.; Hofman, G.L.; Copeland, G.L.

    1983-01-01

    Development and irradiation testing of high-density fuels have been conducted by the US RERTR Program in order to provide the technical means to reduce the enrichment of fuels for research and test reactors. The traditional aluminum dispersion fuel technology has been extended to include the highest practical loadings of uranium-aluminide (UAl/sub x/, 2.3 MgU/m/sup 3/), uranium-oxide (U/sub 3/O/sub 8/, 3.2 MgU/m/sup 3/), and uranium-silicide (U/sub 3/Si/sub 2/, 5.5 MgU/m/sup 3/; U/sub 3/Si, 7.0 MgU/m/sup 3/) fuels. A third uranium-silicide alloy, U/sub 3/SiAl (U + 3.5 wt % Si + 1.5 wt % Al) has been found to perform poorly at high burnup. Testing of miniature fuel plates and full-sized fuel elements is at an advanced stage for the highest loadings of the aluminide and oxide fuels and intermediate loadings of the silicide fuels, and good results have been obtained for low-enriched uranium. The data obtained to date are discussed. 1 reference, 3 figures, 1 table.

  5. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    NASA Astrophysics Data System (ADS)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  6. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  7. EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR

    SciTech Connect

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2011-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  8. High arch

    MedlinePlus

    ... is on the back and balls of the foot (metatarsals head). Your health care provider will check to see if the high ... Call your health care provider if you suspect you are having foot pain related to high arches.

  9. An Innovative High Thermal Conductivity Fuel Design

    SciTech Connect

    PI: James S. Tulenko; Co-PI: Ronald H. Baney,

    2007-10-14

    Uranium dioxide (UO2) is the most common fuel material in commercial nuclear power reactors. UO2 has the advantages of a high melting point, good high-temperature stability, good chemical compatibility with cladding and coolant, and resistance to radiation. The main disadvantage of UO2 is its low thermal conductivity. During a reactor’s operation, because the thermal conductivity of UO2 is very low, for example, about 2.8 W/m-K at 1000 oC [1], there is a large temperature gradient in the UO2 fuel pellet, causing a very high centerline temperature, and introducing thermal stresses, which lead to extensive fuel pellet cracking. These cracks will add to the release of fission product gases after high burnup. The high fuel operating temperature also increases the rate of fission gas release and the fuel pellet swelling caused by fission gases bubbles. The amount of fission gas release and fuel swelling limits the life time of UO2 fuel in reactor. In addition, the high centerline temperature and large temperature gradient in the fuel pellet, leading to a large amount of stored heat, increase the Zircaloy cladding temperature in a lost of coolant accident (LOCA). The rate of Zircaloy-water reaction becomes significant at the temperature above 1200 oC [2]. The ZrO2 layer generated on the surface of the Zircaloy cladding will affect the heat conduction, and will cause a Zircaloy cladding rupture. The objective of this research is to increase the thermal conductivity of UO2, while not affecting the neutronic property of UO2 significantly. The concept to accomplish this goal is to incorporate another material with high thermal conductivity into the UO2 pellet. Silicon carbide (SiC) is a good candidate, because the thermal conductivity of single crystal SiC is 60 times higher than that of UO2 at room temperature and 30 times higher at 800 oC [3]. Silicon carbide also has the properties of low thermal neutron absorption cross section, high melting point, good chemical

  10. Designing CNR, a very high thermal neutron flux facility

    SciTech Connect

    Difilippo, F.C.

    1986-01-01

    According to a recent study (Eastman-Seitz Committee, National Academy of Science) there is a need for a new generation of steady neutron sources with a thermal neutron flux peak between 5 to 10 times 10/sup 15//cm/sup 2/ sec. Ideally the neutron source would have to operate continuously for several days (two weeks at least) with minimum time (2 to 3 days) for refueling and/or maintenance and it would also be used to irradiate materials and produce isotopes. This paper describes the preliminary design of the nuclear reactor for the proposed Center for Neutron Research (CNR). A duplication of existing designs (HFIR, (ORNL), ILL (Grenoble, France)) would imply high total power and small core life; the necessity of higher efficiencies (in terms of peak-flux-per-unit source or power) then becomes apparent. We have found analytical expressions for the efficiency in terms of a few parameters such as the volume of the source and the Fermi age and diffusion length of thermal neutrons in both the source and reflector regions. A single analytical expression can then be used for scoping the design and to intercompare radically different designs. Higher efficiencies can be achieved by reducing the volume and the moderation of a core immersed in a very low absorbing reflector; on the contrary a very long core life has a negative effect on the efficiency at beginning of life. Consequently, and after detailed calculations, we have found a candidate design with the following characteristics: core, U/sub 3/Si/sub 2/, 93% enriched, 18.1-kg /sup 235/U, metal fraction 50%, Al cladding, and 35-L volume; reflector and moderator, D/sub 2/O; efficiency at end of life (EOL) with respect to the ILL reactor, 1.29; flux at EOL, 10 x 10/sup 15//cm/sup 2/ sec (power in core 270. MW); core life, 14 days; burnup 28.4%.

  11. High PRF high current switch

    DOEpatents

    Moran, Stuart L.; Hutcherson, R. Kenneth

    1990-03-27

    A triggerable, high voltage, high current, spark gap switch for use in pu power systems. The device comprises a pair of electrodes in a high pressure hydrogen environment that is triggered by introducing an arc between one electrode and a trigger pin. Unusually high repetition rates may be obtained by undervolting the switch, i.e., operating the trigger at voltages much below the self-breakdown voltage of the device.

  12. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    SciTech Connect

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  13. High arch

    MedlinePlus

    Pes cavus; High foot arch ... High foot arches are much less common than flat feet. They are more likely to be caused by a ... stress is placed on the section of the foot between the ankle and toes (metatarsals). This condition ...

  14. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    SciTech Connect

    Michael A. Pope

    2011-10-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  15. High Stakes

    ERIC Educational Resources Information Center

    Elgin, Catherine Z.

    2004-01-01

    I discuss the contributions of Harvey Siegel, Francis Schrag and Randall Curren to this volume. Their articles cast in bold relief the relation of High Stakes Testing to the goals of education, the nature of mind and the demands of justice. I argue that the connections are deep, but that the considerations these authors raise do not show that High…

  16. High Class

    ERIC Educational Resources Information Center

    Waldecker, Mark

    2005-01-01

    Education administrators make buying decisions for furniture based on many factors. Cost, durability, functionality, safety and aesthetics represent just a few. Those issues always will be important, but gaining greater recognition in recent years has been the role furniture plays in creating positive, high-performance learning environments. The…

  17. Flying High

    ERIC Educational Resources Information Center

    Litvin, Margaret

    2000-01-01

    Nearly a decade ago, the Montreal School of Aerospace Trades became part of a massive education reform that takes most career education in the province out of comprehensive high schools and invests millions of dollars in specialized vocational schools. Through education-industry partnerships, these schools helping to change the image of vocational…

  18. Highly dominating, highly authoritarian personalities.

    PubMed

    Altemeyer, Bob

    2004-08-01

    The author considered the small part of the population whose members score highly on both the Social Dominance Orientation scale and the Right-Wing Authoritarianism scale. Studies of these High SDO-High RWAs, culled from samples of nearly 4000 Canadian university students and over 2600 of their parents and reported in the present article, reveal that these dominating authoritarians are among the most prejudiced persons in society. Furthermore, they seem to combine the worst elements of each kind of personality, being power-hungry, unsupportive of equality, manipulative, and amoral, as social dominators are in general, while also being religiously ethnocentric and dogmatic, as right-wing authoritarians tend to be. The author suggested that, although they are small in number, such persons can have considerable impact on society because they are well-positioned to become the leaders of prejudiced right-wing political movements. PMID:15279331

  19. Creating NDA working standards through high-fidelity spent fuel modeling

    SciTech Connect

    Skutnik, Steven E; Gauld, Ian C; Romano, Catherine E; Trellue, Holly

    2012-01-01

    The Next Generation Safeguards Initiative (NGSI) is developing advanced non-destructive assay (NDA) techniques for spent nuclear fuel assemblies to advance the state-of-the-art in safeguards measurements. These measurements aim beyond the capabilities of existing methods to include the evaluation of plutonium and fissile material inventory, independent of operator declarations. Testing and evaluation of advanced NDA performance will require reference assemblies with well-characterized compositions to serve as working standards against which the NDA methods can be benchmarked and for uncertainty quantification. To support the development of standards for the NGSI spent fuel NDA project, high-fidelity modeling of irradiated fuel assemblies is being performed to characterize fuel compositions and radiation emission data. The assembly depletion simulations apply detailed operating history information and core simulation data as it is available to perform high fidelity axial and pin-by-pin fuel characterization for more than 1600 nuclides. The resulting pin-by-pin isotopic inventories are used to optimize the NDA measurements and provide information necessary to unfold and interpret the measurement data, e.g., passive gamma emitters, neutron emitters, neutron absorbers, and fissile content. A key requirement of this study is the analysis of uncertainties associated with the calculated compositions and signatures for the standard assemblies; uncertainties introduced by the calculation methods, nuclear data, and operating information. An integral part of this assessment involves the application of experimental data from destructive radiochemical assay to assess the uncertainty and bias in computed inventories, the impact of parameters such as assembly burnup gradients and burnable poisons, and the influence of neighboring assemblies on periphery rods. This paper will present the results of high fidelity assembly depletion modeling and uncertainty analysis from independent

  20. Development of high-fidelity multiphysics system for light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Magedanz, Jeffrey W.

    There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining

  1. Irradiation performance of AGR-1 high temperature reactor fuel

    SciTech Connect

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  2. Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

    SciTech Connect

    Eric Shaber; G. Baccaglini; S. Ball; T. Burchell; B. Corwin; T. Fewell; M. Labar; P. MacDonald; P. Rittenhouse; Russ Vollam; F. Southworth

    2003-01-01

    The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.

  3. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

    SciTech Connect

    Fauzia, A. F.; Waris, A.; Novitrian

    2010-06-22

    High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.

  4. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

    NASA Astrophysics Data System (ADS)

    Fauzia, A. F.; Waris, A.; Novitrian

    2010-06-01

    High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950° C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of 235 U enrichment.

  5. High Poverty, High Performing Schools. IDRA Focus.

    ERIC Educational Resources Information Center

    IDRA Newsletter, 1997

    1997-01-01

    This theme issue includes four articles on high performance by poor Texas schools. In "Principal of National Blue Ribbon School Says High Poverty Schools Can Excel" (interview with Robert Zarate by Christie L. Goodman), the principal of Mary Hull Elementary School (San Antonio, Texas) describes how the high-poverty, high-minority school…

  6. Nuclear safety. Technical progress journal, October 1996--December 1996

    SciTech Connect

    1996-12-31

    The five papers in this issue address various issues associated with the behavior of high burnup fuels, especially under reactivity initiated accident (RIA) conditions. The mechanisms and parameters that have an effect on the fuel behavior are detailed, based on tests and analyses. The ultimate goal of the research reported is the development of new regulatory criteria for high burnup fuel under design basis accident conditions. Specific topics of the papers, which are abstracted individually in the database, are: (1) regulatory assessment of test data for RIAs, (2) high burnup fuel transient behavior under RIA conditions, (3) NSRR/RIA experiments with high burnup PWR fuels, (4) the Russian RIA research program, and (5) RIA simulation experiments on the intermediate and high burnup test rods. The papers are contributed from the United States, France, Japan, and Russia.

  7. Fuel Safety Activities in Korea

    SciTech Connect

    Auh, Geun-Sun; Shin, A.D.; Lee, J.S.; Woo, S.W.; Ryu, Y.H.; Kim, Jun-Hwan; Kim, S.K.; Jeong, Y.H.

    2007-07-01

    The current regulatory requirements for fuel performance were based on earlier test data of fresh or low burnup Zircaloy fuels of less than 40 GWD/MTU. Most countries have not changed the current regulatory requirements even if they are actively investigating the high burnup and new cladding alloy effects. Korea agrees with commonly accepted international consensus that although there are technical issues requiring resolutions, these issues do not constitute immediate safety concerns. The high burnup fuel reactor performance experiences of Korea do not show any major problems even if there have been some burnup related fuel failures which are described in the paper. KINS has recommended the industry to have lower fuel failure rates than 1-2 per 50,000 fuel rods. A research project of High Burnup Fuel Safety Tests and Evaluations has started in 2002 under a joint cooperation of KAERI/KNFC/KEPRI and KINS to obtain performance results of high burnup fuel and to develop evaluation technologies of high burnup fuel safety issues. From 1998, KINS has closely monitored and actively participated in international activities such as OECD/NEA CABRI Water Loop Program to reflect on regulatory requirements if needed. KINS will closely monitor the high burnup fuel performances of Korea to strength the regulatory activities if needed. The research activities in Korea including of LOCA and RIA being performed at KAERI with active supports of the industry are summarized in the paper. (authors)

  8. Project Georgia High School/High Tech

    NASA Technical Reports Server (NTRS)

    2000-01-01

    Georgia High School/High Tech has been developing a suggested curriculum for use in its programs. The purpose of this instructional material is to provide a basic curriculum format for teachers of High School/High Tech students. The curriculum is designed to implement QCC classroom instruction that encourages career development in technological fields through post-secondary education, paid summer internships, and exposure to experiences in high technology.

  9. High Blood Pressure (Hypertension)

    MedlinePlus

    ... Print Page Text Size: A A A Listen High Blood Pressure (Hypertension) Nearly 1 in 3 American adults has high ... weight. How Will I Know if I Have High Blood Pressure? High blood pressure is a silent problem — you ...

  10. The Complete Burning of Weapons Grade Plutonium and Highly Enriched Uranium with (Laser Inertial Fusion-Fission Energy) LIFE Engine

    SciTech Connect

    Farmer, J C; Diaz de la Rubia, T; Moses, E

    2008-12-23

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission blanket in a fusion

  11. Highly Anisotropic, Highly Transparent Wood Composites.

    PubMed

    Zhu, Mingwei; Song, Jianwei; Li, Tian; Gong, Amy; Wang, Yanbin; Dai, Jiaqi; Yao, Yonggang; Luo, Wei; Henderson, Doug; Hu, Liangbing

    2016-07-01

    For the first time, two types of highly anisotropic, highly transparent wood composites are demonstrated by taking advantage of the macro-structures in original wood. These wood composites are highly transparent with a total transmittance up to 90% but exhibit dramatically different optical and mechanical properties. PMID:27147136

  12. Project Georgia High School/High Tech

    NASA Technical Reports Server (NTRS)

    2000-01-01

    The High School/High Tech initiative of the President's Committee on Employment of Disabilities, Georgia's application of the collaborative "Georgia Model" and NASA's commitment of funding have shown that opportunities for High School/High Tech students are unlimited. In Georgia, the partnership approach to meeting the needs of this program has opened doors previously closed. As the program grows and develops, reflecting the needs of our students and the marketplace, more opportunities will be available. Our collaboratives are there to provide these opportunities and meet the challenge of matching our students with appropriate education and career goals. Summing up the activities and outcomes of Project Georgia High School/High Tech is not difficult. Significant outcomes have already occurred in the Savannah area as a result of NASA's grant. The support of NASA has enabled Georgia Committee to "grow" High School/High Tech throughout the region-and, by example, the state. The success of the Columbus pilot project has fostered the proliferation of projects, resulting in more than 30 Georgia High School High Tech programs-with eight in the Savannah area.

  13. Postirradiation examination of high-U-loaded low-enriched U/sub 3/O/sub 8/, UAl/sub 2/, and U/sub 3/Si test fuel plates

    SciTech Connect

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofmann, G.; Snelgrove, J.L.

    1984-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded low-enriched U/sub 3/O/sub 8/, UAl/sub 2/ and U/sub 3/Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examinated, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory poerformance for the oxides, aluminides and silicides (except for the highest-loaded U/sub 3/Si plate) with the only indication of detrimental behavior during the slight bowing of some plates at about 80% burnup.

  14. High Blood Pressure

    MedlinePlus

    ... version of this page please turn Javascript on. High Blood Pressure What Is High Blood Pressure? High blood pressure is a common disease in ... the heart, kidneys, brain, and eyes. Types of High Blood Pressure There are two main types of high blood ...

  15. A high-fidelity Monte Carlo evaluation of CANDU-6 safety parameters

    SciTech Connect

    Kim, Y.; Hartanto, D.

    2012-07-01

    Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANDU-6 (CANada Deuterium Uranium) reactor have been evaluated by using a modified MCNPX code. For accurate analysis of the parameters, the DBRC (Doppler Broadening Rejection Correction) scheme was implemented in MCNPX in order to account for the thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted by using the MCNPX and the FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated by using several cross section libraries such as ENDF/B-VI, ENDF/B-VII, JEFF, JENDLE. The PCR value is also evaluated at mid-burnup conditions to characterize safety features of equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, huge number of neutron histories are considered in this work and the standard deviation of the k-inf values is only 0.5{approx}1 pcm. It has been found that the FTC is significantly enhanced by accounting for the Doppler broadening of scattering resonance and the PCR are clearly improved. (authors)

  16. Kr and Xe irradiations in lanthanum (La) doped ceria: Study at the high dose regime

    NASA Astrophysics Data System (ADS)

    Yun, Di; Oaks, Aaron J.; Chen, Wei-ying; Kirk, Marquis A.; Rest, Jeffrey; Insopov, Zinetula Z.; Yacout, Abdellatif M.; Stubbins, James F.

    2011-11-01

    In order to understand cavity and bubble formation and growth in oxide nuclear fuel materials, ion beam irradiation experiments were conducted with two common fission gas species: Kr and Xe. Ceria (CeO 2) was selected as a surrogate material for uranium dioxide (UO 2) due to its many similar properties to UO 2. Ion beam energies were chosen such that both cavities and gas bubbles structures were induced by ion irradiations. The ion irradiation experiments were carried out at 600 °C, at which temperature, cavity/gas bubble structures are believed to be immobile in this material. Lanthanum (La) was chosen as a dopant in CeO 2 to investigate the effect of impurities. The presence of La in the CeO 2 lattice also introduces a predictable initial concentration of oxygen vacancies, similar to the introduction of oxygen vacancies by the existence of Pu 3+ in MOX fuel [1]. The influence of two La concentrations, 5% and 25%, were examined. The study focused on the high dose regime where cavity/gas bubble structures were clearly identifiable with their sizes and number densities readily measurable. Cavity/gas bubble coarsening by coalescence was identified with TEM (Transmission Electron Microscopy) characterizations of as-irradiated La doped CeO 2 specimens. The results revealed that lanthanum trapping has significant influence on the cavity/bubble growth in the material lattice by comparing the cavity/gas bubble size distributions between 5% La doped ceria and 25% La doped ceria. Lattice and kinetic Monte Carlo calculations described in a previous work have provided insights to the interpretations of the experimental results [2]. Solid state Xe precipitates were observed in low energy Xe implantation in 5% La doped ceria to a very high fluence of 1 × 10 17 ions/cm 2 at 600 °C. The solid state Xe precipitate structures are represented by faceted morphology. Very similar observations of solid state/near solid state Xe bubbles were made by Nogita et al. in the outer region

  17. High mechanical performance of Areva upgraded fuel assemblies for PWR in USA

    SciTech Connect

    Gottuso, Dennis; Canat, Jean-Noel; Mollard, Pierre

    2007-07-01

    maintain their performance in all operating environments, in high duty conditions and at extended burnups. AREVA's new family of upgraded products is offered to our customers worldwide. After first LFA's (Lead Fuel Assemblies) in Europe in 2006, and in the US in 2007, a first reload in a European 15x15 reactor will be loaded in 2007. This supply of upgraded products to our customers fully supports the more demanding requirements requested by utilities. The well proven characteristics of all components and their combination bring proven robustness to the products of AREVA's new family of fuel assemblies. (authors)

  18. High blood sugar

    MedlinePlus

    ... page: //medlineplus.gov/ency/patientinstructions/000332.htm High blood sugar To use the sharing features on this page, ... later when energy is needed. Symptoms of High Blood Sugar Symptoms of high blood sugar can include: Being ...

  19. High Blood Pressure

    MedlinePlus

    ... version High Blood Pressure Overview What is blood pressure? Blood pressure is the amount of force that your ... called your blood pressure. What is high blood pressure? High blood pressure (also called hypertension) occurs when your blood ...

  20. High blood pressure - infants

    MedlinePlus

    National High Blood Pressure Education Program Working Group on High Blood Pressure in Children and Adolescents. The fourth report on the diagnosis, evaluation, and treatment of high blood pressure in children and adolescents. Pediatrics . ...

  1. High Blood Pressure

    MedlinePlus

    ... normal blood pressure 140/90 or higher is high blood pressure Between 120 and 139 for the top number, ... prehypertension. Prehypertension means you may end up with high blood pressure, unless you take steps to prevent it. High ...

  2. High-Flying High-Poverty Schools

    ERIC Educational Resources Information Center

    American Educator, 2013

    2013-01-01

    In discussing socioeconomic integration before audiences, the author is frequently asked: What about high-poverty schools that do work? Don't they suggest that economic segregation isn't much of a problem after all? High-poverty public schools that beat the odds paint a heartening story that often attracts considerable media attention. In 2000,…

  3. High Performance Networks for High Impact Science

    SciTech Connect

    Scott, Mary A.; Bair, Raymond A.

    2003-02-13

    This workshop was the first major activity in developing a strategic plan for high-performance networking in the Office of Science. Held August 13 through 15, 2002, it brought together a selection of end users, especially representing the emerging, high-visibility initiatives, and network visionaries to identify opportunities and begin defining the path forward.

  4. High strength high modulus ceramic fiber

    NASA Technical Reports Server (NTRS)

    Fetterolf, R. N.

    1972-01-01

    Low cost method was developed for producing high strength, high modulus, continuous ceramic oxide fibers. Process transforms inexpensive metallic salts into syrup-like liquids that can be fiberized at room temperatures. Resulting salt fibers are then converted to oxides by calcination at relatively low temperatures.

  5. Development of high strength, high temperature ceramics

    NASA Technical Reports Server (NTRS)

    Hall, W. B.

    1982-01-01

    Improvement in the high-pressure turbopumps, both fuel and oxidizer, in the Space Shuttle main engine were considered. The operation of these pumps is limited by temperature restrictions of the metallic components used in these pumps. Ceramic materials that retain strength at high temperatures and appear to be promising candidates for use as turbine blades and impellers are discussed. These high strength materials are sensitive to many related processing parameters such as impurities, sintering aids, reaction aids, particle size, processing temperature, and post thermal treatment. The specific objectives of the study were to: (1) identify and define the processing parameters that affect the properties of Si3N4 ceramic materials, (2) design and assembly equipment required for processing high strength ceramics, (3) design and assemble test apparatus for evaluating the high temperature properties of Si3N4, and (4) conduct a research program of manufacturing and evaluating Si3N4 materials as applicable to rocket engine applications.

  6. High pressure, high current, low inductance, high reliability sealed terminals

    DOEpatents

    Hsu, John S [Oak Ridge, TN; McKeever, John W [Oak Ridge, TN

    2010-03-23

    The invention is a terminal assembly having a casing with at least one delivery tapered-cone conductor and at least one return tapered-cone conductor routed there-through. The delivery and return tapered-cone conductors are electrically isolated from each other and positioned in the annuluses of ordered concentric cones at an off-normal angle. The tapered cone conductor service can be AC phase conductors and DC link conductors. The center core has at least one service conduit of gate signal leads, diagnostic signal wires, and refrigerant tubing routed there-through. A seal material is in direct contact with the casing inner surface, the tapered-cone conductors, and the service conduits thereby hermetically filling the interstitial space in the casing interior core and center core. The assembly provides simultaneous high-current, high-pressure, low-inductance, and high-reliability service.

  7. High Blood Pressure

    MedlinePlus

    ... page from the NHLBI on Twitter. Description of High Blood Pressure Español High blood pressure is a common disease ... defines high blood pressure severity levels. Stages of High Blood Pressure in Adults Stages Systolic (top number) Diastolic (bottom ...

  8. Bumball: Highly Engaging, Highly Inclusive, and Highly Entertaining

    ERIC Educational Resources Information Center

    Hall, Amber; Barney, David; Wilkinson, Carol

    2014-01-01

    Physical educators are always looking for new and exciting games and activities in which students can participate. This article describes Bumball, a high-intensity game that provides the opportunity for students to use many common game skills, such as hand-eye coordination, passing to a target, running, playing defense, and getting to an open…

  9. High-current, high-frequency capacitors

    NASA Technical Reports Server (NTRS)

    Renz, D. D.

    1983-01-01

    The NASA Lewis high-current, high-frequency capacitor development program was conducted under a contract with Maxwell Laboratories, Inc., San Diego, California. The program was started to develop power components for space power systems. One of the components lacking was a high-power, high-frequency capacitor. Some of the technology developed in this program may be directly usable in an all-electric airplane. The materials used in the capacitor included the following: the film is polypropylene, the impregnant is monoisopropyl biphenyl, the conductive epoxy is Emerson and Cuming Stycast 2850 KT, the foil is aluminum, the case is stainless steel (304), and the electrode is a modified copper-ceramic.

  10. High strength and high toughness steel

    DOEpatents

    Parker, Earl R.; Zackay, Victor F.

    1979-01-01

    A structural steel which possess both high strength and high toughness and has particular application of cryogenic uses. The steel is produced by the utilization of thermally induced phase transformation following heating in a three-phase field in iron-rich alloys of the Fe-Ni-Ti system, with a preferred composition of 12% nickel, 0.5% titanium, the remainder being iron.

  11. High strength, high ductility low carbon steel

    DOEpatents

    Koo, Jayoung; Thomas, Gareth

    1978-01-01

    A high strength, high ductility low carbon steel consisting essentially of iron, 0.05-0.15 wt% carbon, and 1-3 wt% silicon. Minor amounts of other constituents may be present. The steel is characterized by a duplex ferrite-martensite microstructure in a fibrous morphology. The microstructure is developed by heat treatment consisting of initial austenitizing treatment followed by annealing in the (.alpha. + .gamma.) range with intermediate quenching.

  12. High Temperature, High Power Piezoelectric Composite Transducers

    PubMed Central

    Lee, Hyeong Jae; Zhang, Shujun; Bar-Cohen, Yoseph; Sherrit, StewarT.

    2014-01-01

    Piezoelectric composites are a class of functional materials consisting of piezoelectric active materials and non-piezoelectric passive polymers, mechanically attached together to form different connectivities. These composites have several advantages compared to conventional piezoelectric ceramics and polymers, including improved electromechanical properties, mechanical flexibility and the ability to tailor properties by using several different connectivity patterns. These advantages have led to the improvement of overall transducer performance, such as transducer sensitivity and bandwidth, resulting in rapid implementation of piezoelectric composites in medical imaging ultrasounds and other acoustic transducers. Recently, new piezoelectric composite transducers have been developed with optimized composite components that have improved thermal stability and mechanical quality factors, making them promising candidates for high temperature, high power transducer applications, such as therapeutic ultrasound, high power ultrasonic wirebonding, high temperature non-destructive testing, and downhole energy harvesting. This paper will present recent developments of piezoelectric composite technology for high temperature and high power applications. The concerns and limitations of using piezoelectric composites will also be discussed, and the expected future research directions will be outlined. PMID:25111242

  13. High temperature, high power piezoelectric composite transducers.

    PubMed

    Lee, Hyeong Jae; Zhang, Shujun; Bar-Cohen, Yoseph; Sherrit, Stewart

    2014-01-01

    Piezoelectric composites are a class of functional materials consisting of piezoelectric active materials and non-piezoelectric passive polymers, mechanically attached together to form different connectivities. These composites have several advantages compared to conventional piezoelectric ceramics and polymers, including improved electromechanical properties, mechanical flexibility and the ability to tailor properties by using several different connectivity patterns. These advantages have led to the improvement of overall transducer performance, such as transducer sensitivity and bandwidth, resulting in rapid implementation of piezoelectric composites in medical imaging ultrasounds and other acoustic transducers. Recently, new piezoelectric composite transducers have been developed with optimized composite components that have improved thermal stability and mechanical quality factors, making them promising candidates for high temperature, high power transducer applications, such as therapeutic ultrasound, high power ultrasonic wirebonding, high temperature non-destructive testing, and downhole energy harvesting. This paper will present recent developments of piezoelectric composite technology for high temperature and high power applications. The concerns and limitations of using piezoelectric composites will also be discussed, and the expected future research directions will be outlined. PMID:25111242

  14. High-temperature-staged fluidized-bed combustion (HITS), bench scale experimental test program conducted during 1980. Final report

    SciTech Connect

    Anderson, R E; Jassowski, D M; Newton, R A; Rudnicki, M L

    1981-04-01

    An experimental program was conducted to evaluate the process feasibility of the first stage of the HITS two-stage coal combustion system. Tests were run in a small (12-in. ID) fluidized bed facility at the Energy Engineering Laboratory, Aerojet Energy Conversion Company, Sacramento, California. The first stage reactor was run with low (0.70%) and high (4.06%) sulfur coals with ash fusion temperatures of 2450/sup 0/ and 2220/sup 0/F, respectively. Limestone was used to scavenge the sulfur. The produced low-Btu gas was burned in a combustor. Bed temperature and inlet gas percent oxygen were varied in the course of testing. Key results are summarized as follows: the process was stable and readily controllable, and generated a free-flowing char product using coals with low (2220/sup 0/F) and high (2450/sup 0/F) ash fusion temperatures at bed temperatures of at least 1700/sup 0/ and 1800/sup 0/F, respectively; the gaseous product was found to have a total heating value of about 120 Btu/SCF at 1350/sup 0/F, and the practicality of cleaning the hot product gas and delivering it to the combustor was demonstrated; sulfur capture efficiencies above 80% were demonstrated for both low and high sulfur coals with a calcium/sulfur mole ratio of approximately two; gasification rates of about 5,000 SCF/ft/sup 2/-hr were obtained for coal input rates ranging from 40 to 135 lbm/hr, as required to maintain the desired bed temperatures; and the gaseous product yielded combustion temperatures in excess of 3000/sup 0/F when burned with preheated (900/sup 0/F) air. The above test results support the promise of the HITS system to provide a practical means of converting high sulfur coal to a clean gas for industrial applications. Sulfur capture, gas heating value, and gas production rate are all in the range required for an effective system. Planning is underway for additional testing of the system in the 12-in. fluid bed facility, including demonstration of the second stage char burnup

  15. High assurance SPIRAL

    NASA Astrophysics Data System (ADS)

    Franchetti, Franz; Sandryhaila, Aliaksei; Johnson, Jeremy R.

    2014-06-01

    In this paper we introduce High Assurance SPIRAL to solve the last mile problem for the synthesis of high assurance implementations of controllers for vehicular systems that are executed in today's and future embedded and high performance embedded system processors. High Assurance SPIRAL is a scalable methodology to translate a high level specification of a high assurance controller into a highly resource-efficient, platform-adapted, verified control software implementation for a given platform in a language like C or C++. High Assurance SPIRAL proves that the implementation is equivalent to the specification written in the control engineer's domain language. Our approach scales to problems involving floating-point calculations and provides highly optimized synthesized code. It is possible to estimate the available headroom to enable assurance/performance trade-offs under real-time constraints, and enables the synthesis of multiple implementation variants to make attacks harder. At the core of High Assurance SPIRAL is the Hybrid Control Operator Language (HCOL) that leverages advanced mathematical constructs expressing the controller specification to provide high quality translation capabilities. Combined with a verified/certified compiler, High Assurance SPIRAL provides a comprehensive complete solution to the efficient synthesis of verifiable high assurance controllers. We demonstrate High Assurance SPIRALs capability by co-synthesizing proofs and implementations for attack detection and sensor spoofing algorithms and deploy the code as ROS nodes on the Landshark unmanned ground vehicle and on a Synthetic Car in a real-time simulator.

  16. High blood pressure

    MedlinePlus

    ... this page: //medlineplus.gov/ency/article/000468.htm High blood pressure To use the sharing features on ... body. Hypertension is the term used to describe high blood pressure. Blood pressure readings are given as ...

  17. High Speed data acquisition

    SciTech Connect

    Cooper, Peter S.

    1998-02-01

    A general introduction to high Speed data acquisition system techniques in modern particle physics experiments is given. Examples are drawn from the SELEX(E781) high statistics charmed baryon production and decay experiment now taking data at Fermilab.

  18. High-Altitude Illness

    MedlinePlus

    ... altitude illness: Acute mountain sickness High-altitude pulmonary edema (also called HAPE), which affects the lungs High-altitude cerebral edema (also called HACE), which affects the brain These ...

  19. Cleaning Up High.

    ERIC Educational Resources Information Center

    Rittner-Heir, Robbin M.

    2001-01-01

    Discusses cleaning techniques and equipment for areas in a school that are hard to reach, such as the high spacious ceilings, the tall windows of a school atrium, and high points in gymnasiums and auditoriums. (GR)

  20. High blood pressure - infants

    MedlinePlus

    Hypertension - infants ... and blood vessels The health of the kidneys High blood pressure in infants may be due to kidney or ... Bronchopulmonary dysplasia Renal artery stenosis In newborn babies, high blood pressure is often caused by a blood clot in ...

  1. High potassium level

    MedlinePlus

    High potassium level is a problem in which the amount of potassium in the blood is higher than normal. The medical ... There are often no symptoms with a high level of potassium. When symptoms do occur, they may ...

  2. High energy neutron radiography

    SciTech Connect

    Gavron, A.; Morley, K.; Morris, C.; Seestrom, S.; Ullmann, J.; Yates, G.; Zumbro, J.

    1996-06-01

    High-energy spallation neutron sources are now being considered in the US and elsewhere as a replacement for neutron beams produced by reactors. High-energy and high intensity neutron beams, produced by unmoderated spallation sources, open potential new vistas of neutron radiography. The authors discuss the basic advantages and disadvantages of high-energy neutron radiography, and consider some experimental results obtained at the Weapons Neutron Research (WNR) facility at Los Alamos.

  3. Tunable high pressure lasers

    NASA Technical Reports Server (NTRS)

    Hess, R. V.

    1976-01-01

    Atmospheric transmission of high energy CO2 lasers is considerably improved by high pressure operation which, due to pressure broadening, permits tuning the laser lines off atmospheric absorption lines. Pronounced improvement is shown for horizontal transmission at altitudes above several kilometers and for vertical transmission through the entire atmosphere. Applications of tunable high pressure CO2 lasers to energy transmission and to remote sensing are discussed along with initial efforts in tuning high pressure CO2 lasers.

  4. High power microwave generator

    DOEpatents

    Ekdahl, C.A.

    1983-12-29

    A microwave generator efficiently converts the energy of an intense relativistic electron beam (REB) into a high-power microwave emission using the Smith-Purcell effect which is related to Cerenkov radiation. Feedback for efficient beam bunching and high gain is obtained by placing a cylindrical Smith-Purcell transmission grating on the axis of a toroidal resonator. High efficiency results from the use of a thin cold annular highly-magnetized REB that is closely coupled to the resonant structure.

  5. High power microwave generator

    DOEpatents

    Ekdahl, Carl A.

    1986-01-01

    A microwave generator efficiently converts the energy of an intense relativistic electron beam (REB) into a high-power microwave emission using the Smith-Purcell effect which is related to Cerenkov radiation. Feedback for efficient beam bunching and high gain is obtained by placing a cylindrical Smith-Purcell transmission grating on the axis of a toroidal resonator. High efficiency results from the use of a thin cold annular highly-magnetized REB that is closely coupled to the resonant structure.

  6. An Innovative High Thermal Conductivity Fuel Design

    SciTech Connect

    Jamil A. Khan

    2009-11-21

    Thermal conductivity of the fuel in today's Light Water Reactors, Uranium dioxide, can be improved by incorporating a uniformly distributed heat conducting network of a higher conductivity material, Silicon Carbide. The higher thermal conductivity of SiC along with its other prominent reactor-grade properties makes it a potential material to address some of the related issues when used in UO2 [97% TD]. This ongoing research, in collaboration with the University of Florida, aims to investigate the feasibility and develop a formal methodology of producing the resultant composite oxide fuel. Calculations of effective thermal conductivity of the new fuel as a function of %SiC for certain percentages and as a function of temperature are presented as a preliminary approach. The effective thermal conductivities are obtained at different temperatures from 600K to 1600K. The corresponding polynomial equations for the temperature-dependent thermal conductivities are given based on the simulation results. Heat transfer mechanism in this fuel is explained using a finite volume approach and validated against existing empirical models. FLUENT 6.1.22 was used for thermal conductivity calculations and to estimate reduction in centerline temperatures achievable within such a fuel rod. Later, computer codes COMBINE-PC and VENTURE-PC were deployed to estimate the fuel enrichment required, to maintain the same burnup levels, corresponding to a volume percent addition of SiC.

  7. High Technology Partnership Project.

    ERIC Educational Resources Information Center

    Francis Tuttle Vo-Tech Center, Oklahoma City, OK.

    The High Technology Center at Francis Tuttle Vo-Tech Center in Oklahoma City conducted an 18-month demonstration program, beginning in January 1989, to train or retrain average workers, unemployed because of cutbacks in their field or lack of marketable skills, for careers in high technology. The High Technology Center offered adults training in…

  8. High Technology Needs Assessment.

    ERIC Educational Resources Information Center

    Southeastern Louisiana Univ., Hammond.

    A project produced a high technology status report providing needs assessment data for educational planning. The purpose was to determine the impact and future of high technology in Louisiana. Information was obtained from 68 Louisiana manufacturing industries by mailed questionnaire. Data indicated that 45 industries were involved in high tech. A…

  9. High performance systems

    SciTech Connect

    Vigil, M.B.

    1995-03-01

    This document provides a written compilation of the presentations and viewgraphs from the 1994 Conference on High Speed Computing given at the High Speed Computing Conference, {open_quotes}High Performance Systems,{close_quotes} held at Gleneden Beach, Oregon, on April 18 through 21, 1994.

  10. High speed civil transport

    NASA Technical Reports Server (NTRS)

    Mcknight, R. L.

    1992-01-01

    The design requirements of the High Speed Civil Transport (HSCT) are discussed. The following design concerns are presented: (1) environmental impact (emissions and noise); (2) critical components (the high temperature combustor and the lightweight exhaust nozzle); and (3) advanced materials (high temperature ceramic matrix composites (CMC's)/intermetallic matrix composites (IMC's)/metal matrix composites (MMC's)).

  11. Early College High Schools

    ERIC Educational Resources Information Center

    Dessoff, Alan

    2011-01-01

    For at-risk students who stand little chance of going to college, or even finishing high school, a growing number of districts have found a solution: Give them an early start in college while they still are in high school. The early college high school (ECHS) movement that began with funding from the Bill and Melinda Gates Foundation 10 years ago…

  12. Evaluating High School IT

    ERIC Educational Resources Information Center

    Thompson, Brett A.

    2004-01-01

    Since its inception in 1997, Cisco's curriculum has entered thousands of high schools across the U.S. and around the world for two reasons: (1) Cisco has a large portion of the computer networking market, and thus has the resources for and interest in developing high school academies; and (2) high school curriculum development teams recognize the…

  13. High Tech High: Cruising on the Internet.

    ERIC Educational Resources Information Center

    Wilson, Thomas F.

    1995-01-01

    A "driver's license" at one suburban Minnesota high school is a free round-trip ticket to the world. The Internet allows students to stroll through the Louvre, tour the White House, or examine current satellite weather photos at NASA's computer center. Students are developing friendships, exploring different cultures, and learning foreign…

  14. High power, high beam quality regenerative amplifier

    DOEpatents

    Hackel, Lloyd A.; Dane, Clifford B.

    1993-01-01

    A regenerative laser amplifier system generates high peak power and high energy per pulse output beams enabling generation of X-rays used in X-ray lithography for manufacturing integrated circuits. The laser amplifier includes a ring shaped optical path with a limited number of components including a polarizer, a passive 90 degree phase rotator, a plurality of mirrors, a relay telescope, and a gain medium, the components being placed close to the image plane of the relay telescope to reduce diffraction or phase perturbations in order to limit high peak intensity spiking. In the ring, the beam makes two passes through the gain medium for each transit of the optical path to increase the amplifier gain to loss ratio. A beam input into the ring makes two passes around the ring, is diverted into an SBS phase conjugator and proceeds out of the SBS phase conjugator back through the ring in an equal but opposite direction for two passes, further reducing phase perturbations. A master oscillator inputs the beam through an isolation cell (Faraday or Pockels) which transmits the beam into the ring without polarization rotation. The isolation cell rotates polarization only in beams proceeding out of the ring to direct the beams out of the amplifier. The diffraction limited quality of the input beam is preserved in the amplifier so that a high power output beam having nearly the same diffraction limited quality is produced.

  15. High power, high beam quality regenerative amplifier

    DOEpatents

    Hackel, L.A.; Dane, C.B.

    1993-08-24

    A regenerative laser amplifier system generates high peak power and high energy per pulse output beams enabling generation of X-rays used in X-ray lithography for manufacturing integrated circuits. The laser amplifier includes a ring shaped optical path with a limited number of components including a polarizer, a passive 90 degree phase rotator, a plurality of mirrors, a relay telescope, and a gain medium, the components being placed close to the image plane of the relay telescope to reduce diffraction or phase perturbations in order to limit high peak intensity spiking. In the ring, the beam makes two passes through the gain medium for each transit of the optical path to increase the amplifier gain to loss ratio. A beam input into the ring makes two passes around the ring, is diverted into an SBS phase conjugator and proceeds out of the SBS phase conjugator back through the ring in an equal but opposite direction for two passes, further reducing phase perturbations. A master oscillator inputs the beam through an isolation cell (Faraday or Pockels) which transmits the beam into the ring without polarization rotation. The isolation cell rotates polarization only in beams proceeding out of the ring to direct the beams out of the amplifier. The diffraction limited quality of the input beam is preserved in the amplifier so that a high power output beam having nearly the same diffraction limited quality is produced.

  16. Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

    SciTech Connect

    P. D. Wheatley; R. P. Rechard

    1998-09-01

    The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the U.S. Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF). The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW). A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP). A performance assessment (PA) was chosen as the method of analysis. The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr. Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis. Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria. This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (1.5%) and mass (3.8%) of waste in the potential repository. Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF.

  17. Burnup Predictions for Metal Fuel Tests in the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Nelson, Joseph V.

    2012-06-01

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The FFTF operated successfully from initial startup in 1980 through the end of the last operating cycle in March, 1992. A variety of fuel tests were irradiated in FFTF to provide performance data over a range of conditions. The MFF-3 and MFF-5 tests were U10Zr metal fuel tests with HT9 cladding. The MFF-3 and MFF-5 tests were both aggressive irradiation tests of U10Zr metal fuel pins with HT9 cladding that were prototypic of full scale LMR designs. MFF-3 was irradiated for 726 Effective Full Power Days (EFPD), starting from Cycle 10C1 (from November 1988 through March 1992), and MFF-5 was irradiated for 503 EFPD starting from Cycle 11B1 (from January 1990 through March 1992). A group of fuel pins from these two tests are undergoing post irradiation examination at the Idaho National Laboratory (INL) for the Fuel Cycle Research and Development Program (FCRD). The generation of a data package of key information on the irradiation environment and current pin detailed compositions for these tests is described. This information will be used in interpreting the results of these examinations.

  18. Design and burn-up analyses of new type holder for silicon neutron transmutation doping.

    PubMed

    Komeda, Masao; Arai, Masaji; Tamai, Kazuo; Kawasaki, Kozo

    2016-07-01

    We have developed a new silicon irradiation holder with a neutron filter to increase the irradiation efficiency. The neutron filter is made of an alloy of aluminum and B4C particles. We fabricated a new holder based on the results of design analyses. This filter has limited use in applications requiring prolonged use due to a decrease in the amount of (10)B in B4C particles. We investigated the influence of (10)B reduction on doping distribution in a silicon ingot by using the Monte Carlo Code MVP. PMID:27131643

  19. Determination of Light Water Reactor Fuel Burnup with the Isotope Ratio Method

    SciTech Connect

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2007-11-01

    For the current project to demonstrate that isotope ratio measurements can be extended to zirconium alloys used in LWR fuel assemblies we report new analyses on irradiated samples obtained from a reactor. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers new measurements done on irradiated and unirradiated zirconium alloys, Unirradiated zircaloy samples serve as reference samples and indicate starting values or natural values for the Ti isotope ratio measured. New measurements of irradiated samples include results for 3 samples provided by AREVA. New results indicate: 1. Titanium isotope ratios were measured again in unirradiated samples to obtain reference or starting values at the same time irradiated samples were analyzed. In particular, 49Ti/48Ti ratios were indistinguishably close to values determined several months earlier and to expected natural values. 2. 49Ti/48Ti ratios were measured in 3 irradiated samples thus far, and demonstrate marked departures from natural or initial ratios, well beyond analytical uncertainty, and the ratios vary with reported fluence values. The irradiated samples appear to have significant surface contamination or radiation damage which required more time for SIMS analyses. 3. Other activated impurity elements still limit the sample size for SIMS analysis of irradiated samples. The sub-samples chosen for SIMS analysis, although smaller than optimal, were still analyzed successfully without violating the conditions of the applicable Radiological Work Permit

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

    SciTech Connect

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

    1995-10-01

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

  1. Moderator poison design and burn-up calculations at the SNS

    NASA Astrophysics Data System (ADS)

    Lu, W.; Ferguson, P. D.; Iverson, E. B.; Gallmeier, F. X.; Popova, I.

    2008-06-01

    The spallation neutron source (SNS) at Oak Ridge National Laboratory was commissioned in April 2006. At the nominal operating power (1.4 MW), it will have thermal neutron fluxes approximately an order of magnitude greater than any existing pulsed spallation source. It thus brings a serious challenge to the lifetime of the moderator poison sheets. The SNS moderators are integrated with the inner reflector plug (IRP) at a cost of ˜$2 million a piece. A replacement of the inner reflector plug presents a significant drawback to the facility due to the activation and the operation cost. Although there are a lot of factors limiting the lifetime of the inner reflector plug, like radiation damage to the structural material and helium production of beryllium, the bottle-neck is the lifetime of the moderator poison sheets. Increasing the thickness of the poison sheet extends the lifetime but would sacrifice the neutronic performance of the moderators. A compromise is accepted at the current SNS target system which uses thick Gd poison sheets at a projected lifetime of 6 MW-years of operation. The calculations in this paper reveal that Cd may be a better poison material from the perspective of lifetime and neutronic performance. In replacing Gd, the inner reflector plug could reach a lifetime of 8 MW-years with ˜5% higher peak neutron fluxes at almost no loss of energy resolution.

  2. Development and validation of ALEPH2 Monte Carlo burn-up code

    SciTech Connect

    Van Den Eynde, G.; Stankovskiy, A.; Fiorito, L.; Broustaut, M.

    2013-07-01

    The ALEPH2 Monte Carlo depletion code has two principal features that make it a flexible and powerful tool for reactor analysis. First of all, its comprehensive nuclear data library ensures the consistency between steady-state Monte Carlo and deterministic depletion modules. It covers neutron and proton induced reactions, neutron and proton fission product yields, spontaneous fission product yields, radioactive decay data and total recoverable energies per fission. Secondly, ALEPH2 uses an advanced numerical solver for the first order ordinary differential equations describing the isotope balances, namely a Radau IIA implicit Runge-Kutta method. The versatility of the code allows using it for time behavior simulation of various systems ranging from single pin model to full-scale reactor model. The code is extensively used for the neutronics design of the MYRRHA research fast spectrum facility which will operate in both critical and sub-critical modes. The code has been validated on the decay heat data from JOYO experimental fast reactor. (authors)

  3. Experimental Investigation of Burnup Credit for Safe Transport, Storage, and Disposal of Spent Nuclear Fuel

    SciTech Connect

    Harms, Gary A.; Helmick, Paul H.; Ford, John T.; Walker, Sharon A.; Berry, Donald T.; Pickard, Paul S.

    2004-04-01

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

  4. Experimental investigation of burnup credit for safe transport, storage, and disposal of spent nuclear fuel.

    SciTech Connect

    Berry, Donald T.; Harms, Gary A.; Ford, John T.; Walker, Sharon Ann; Helmick, Paul H.; Pickard, Paul S.

    2004-04-01

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

  5. High performance polymer development

    NASA Technical Reports Server (NTRS)

    Hergenrother, Paul M.

    1991-01-01

    The term high performance as applied to polymers is generally associated with polymers that operate at high temperatures. High performance is used to describe polymers that perform at temperatures of 177 C or higher. In addition to temperature, other factors obviously influence the performance of polymers such as thermal cycling, stress level, and environmental effects. Some recent developments at NASA Langley in polyimides, poly(arylene ethers), and acetylenic terminated materials are discussed. The high performance/high temperature polymers discussed are representative of the type of work underway at NASA Langley Research Center. Further improvement in these materials as well as the development of new polymers will provide technology to help meet NASA future needs in high performance/high temperature applications. In addition, because of the combination of properties offered by many of these polymers, they should find use in many other applications.

  6. High pressure and high temperature apparatus

    DOEpatents

    Voronov, Oleg A.

    2005-09-13

    A design for high pressure/high temperature apparatus and reaction cell to achieve .about.30 GPa pressure in .about.1 cm volume and .about.100 GPa pressure in .about.1 mm volumes and 20-5000.degree. C. temperatures in a static regime. The device includes profiled anvils (28) action on a reaction cell (14, 16) containing the material (26) to be processed. The reaction cell includes a heater (18) surrounded by insulating layers and screens. Surrounding the anvils are cylindrical inserts and supporting rings (30-48) whose hardness increases towards the reaction cell. These volumes may be increased considerably if applications require it, making use of presses that have larger loading force capability, larger frames and using larger anvils.

  7. High Resolution, High Frame Rate Video Technology

    NASA Technical Reports Server (NTRS)

    1990-01-01

    Papers and working group summaries presented at the High Resolution, High Frame Rate Video (HHV) Workshop are compiled. HHV system is intended for future use on the Space Shuttle and Space Station Freedom. The Workshop was held for the dual purpose of: (1) allowing potential scientific users to assess the utility of the proposed system for monitoring microgravity science experiments; and (2) letting technical experts from industry recommend improvements to the proposed near-term HHV system. The following topics are covered: (1) State of the art in the video system performance; (2) Development plan for the HHV system; (3) Advanced technology for image gathering, coding, and processing; (4) Data compression applied to HHV; (5) Data transmission networks; and (6) Results of the users' requirements survey conducted by NASA.

  8. A simple, high efficiency, high resolution spectropolarimeter

    NASA Astrophysics Data System (ADS)

    Barden, Samuel C.

    2012-09-01

    A simple concept is described that uses volume phase holographic gratings as polarizing dispersers for a high efficiency, high resolution spectropolarimeter. Although the idea has previously been mentioned in the literature as possible, such a concept has not been explored in detail. Performance analysis is presented for a VPHG spectropolarimeter concept that could be utilized for both solar and night-time astronomy. Instrumental peak efficiency can approach 100% with spectral dispersions permitting R~200,000 spectral resolution with diffraction limited telescopes. The instrument has 3-channels: two dispersed image planes with orthogonal polarization and an undispersed image plane. The concept has a range of versatility where it could be configured (with appropriate half-wave plates) for slit-fed spectroscopy or without slits for snapshot/hyperspectral/tomographic spectroscopic imaging. Multiplex gratings could also be used for the simultaneous recording of two separate spectral bands or multiple instruments could be daisy chained with beam splitters for further spectral coverage.

  9. High modulus high temperature glass fibers

    NASA Technical Reports Server (NTRS)

    Bacon, J. F.

    1973-01-01

    The search for a new high-modulus, high-temperature glass fiber involved the preparation of 500 glass compositions lying in 12 glass fields. These systems consisted primarily of low atomic number oxides and rare-earth oxides. Direct optical measurements of the kinetics of crystallization of the cordierite-rare earth system, for example, showed that the addition of rare-earth oxides decreased the rate of formation of cordierite crystals. Glass samples prepared from these systems proved that the rare-earth oxides made large specific contributions to the Young's modulus of the glasses. The best glasses have moduli greater than 21 million psi, the best glass fibers have moduli greater than 18 million psi, and the best glass fiber-epoxy resin composites have tensile strengths of 298,000 psi, compressive strengths of at least 220,000 psi, flexural strengths of 290,000 psi, and short-beam shear strengths of almost 17,000 psi.

  10. High energy high brightness thin disk laser

    NASA Astrophysics Data System (ADS)

    Nixon, Matthew D.; Cates, Michael C.

    2012-11-01

    Boeing has been developing solid state lasers for high energy applications since 2004 using Yb:YAG thin disk lasers as pioneered by A. Giesen1 and commercialized by Trumpf Laser GmbH.2 In this paper, we report results of our second generation design and status of a third generation we are currently developing, which will produce 35 kW and a beam quality <2.

  11. HIGH VOLTAGE, HIGH CURRENT SPARK GAP SWITCH

    DOEpatents

    Dike, R.S.; Lier, D.W.; Schofield, A.E.; Tuck, J.L.

    1962-04-17

    A high voltage and current spark gap switch comprising two main electrodes insulatingly supported in opposed spaced relationship and a middle electrode supported medially between the main electrodes and symmetrically about the median line of the main electrodes is described. The middle electrode has a perforation aligned with the median line and an irradiation electrode insulatingly supported in the body of the middle electrode normal to the median line and protruding into the perforation. (AEC)

  12. High frequency, high power capacitor development

    NASA Technical Reports Server (NTRS)

    White, C. W.; Hoffman, P. S.

    1983-01-01

    A program to develop a special high energy density, high power transfer capacitor to operate at frequency of 40 kHz, 600 V rms at 125 A rms plus 600 V dc bias for space operation. The program included material evaluation and selection, a capacitor design was prepared, a thermal analysis performed on the design. Fifty capacitors were manufactured for testing at 10 kHz and 40 kHz for 50 hours at Industrial Electric Heating Co. of Columbus, Ohio. The vacuum endurance test used on environmental chamber and temperature plate furnished by Maxwell. The capacitors were energized with a special power conditioning apparatus developed by Industrial Electric Heating Co. Temperature conditions of the capacitors were monitored by IEHCo test equipment. Successful completion of the vacuum endurance test series confirmed achievement of the main goal of producing a capacitor or reliable operation at high frequency in an environment normally not hospitable to electrical and electronic components. The capacitor developed compared to a typical commercial capacitor at the 40 kHz level represents a decrease in size and weight by a factor of seven.

  13. High output lamp with high brightness

    DOEpatents

    Kirkpatrick, Douglas A.; Bass, Gary K.; Copsey, Jesse F.; Garber, Jr., William E.; Kwong, Vincent H.; Levin, Izrail; MacLennan, Donald A.; Roy, Robert J.; Steiner, Paul E.; Tsai, Peter; Turner, Brian P.

    2002-01-01

    An ultra bright, low wattage inductively coupled electrodeless aperture lamp is powered by a solid state RF source in the range of several tens to several hundreds of watts at various frequencies in the range of 400 to 900 MHz. Numerous novel lamp circuits and components are disclosed including a wedding ring shaped coil having one axial and one radial lead, a high accuracy capacitor stack, a high thermal conductivity aperture cup and various other aperture bulb configurations, a coaxial capacitor arrangement, and an integrated coil and capacitor assembly. Numerous novel RF circuits are also disclosed including a high power oscillator circuit with reduced complexity resonant pole configuration, parallel RF power FET transistors with soft gate switching, a continuously variable frequency tuning circuit, a six port directional coupler, an impedance switching RF source, and an RF source with controlled frequency-load characteristics. Numerous novel RF control methods are disclosed including controlled adjustment of the operating frequency to find a resonant frequency and reduce reflected RF power, controlled switching of an impedance switched lamp system, active power control and active gate bias control.

  14. Highly directional thermal emitter

    DOEpatents

    Ribaudo, Troy; Shaner, Eric A; Davids, Paul; Peters, David W

    2015-03-24

    A highly directional thermal emitter device comprises a two-dimensional periodic array of heavily doped semiconductor structures on a surface of a substrate. The array provides a highly directional thermal emission at a peak wavelength between 3 and 15 microns when the array is heated. For example, highly doped silicon (HDSi) with a plasma frequency in the mid-wave infrared was used to fabricate nearly perfect absorbing two-dimensional gratings structures that function as highly directional thermal radiators. The absorption and emission characteristics of the HDSi devices possessed a high degree of angular dependence for infrared absorption in the 10-12 micron range, while maintaining high reflectivity of solar radiation (.about.64%) at large incidence angles.

  15. High lift selected concepts

    NASA Technical Reports Server (NTRS)

    Henderson, M. L.

    1979-01-01

    The benefits to high lift system maximum life and, alternatively, to high lift system complexity, of applying analytic design and analysis techniques to the design of high lift sections for flight conditions were determined and two high lift sections were designed to flight conditions. The influence of the high lift section on the sizing and economics of a specific energy efficient transport (EET) was clarified using a computerized sizing technique and an existing advanced airplane design data base. The impact of the best design resulting from the design applications studies on EET sizing and economics were evaluated. Flap technology trade studies, climb and descent studies, and augmented stability studies are included along with a description of the baseline high lift system geometry, a calculation of lift and pitching moment when separation is present, and an inverse boundary layer technique for pressure distribution synthesis and optimization.

  16. High speed handpieces

    PubMed Central

    Bhandary, Nayan; Desai, Asavari; Shetty, Y Bharath

    2014-01-01

    High speed instruments are versatile instruments used by clinicians of all specialties of dentistry. It is important for clinicians to understand the types of high speed handpieces available and the mechanism of working. The centers for disease control and prevention have issued guidelines time and again for disinfection and sterilization of high speed handpieces. This article presents the recent developments in the design of the high speed handpieces. With a view to prevent hospital associated infections significant importance has been given to disinfection, sterilization & maintenance of high speed handpieces. How to cite the article: Bhandary N, Desai A, Shetty YB. High speed handpieces. J Int Oral Health 2014;6(1):130-2. PMID:24653618

  17. High-altitude headache.

    PubMed

    Marmura, Michael J; Hernandez, Pablo Bandres

    2015-05-01

    High-altitude headache is one of many neurological symptoms associated with the ascent to high altitudes. Cellular hypoxia due to decreased barometric pressure seems to be the common final pathway for headache as altitude increases. Susceptibility to high-altitude headache depends on genetic factors, history of migraine, and acclimatization, but symptoms of acute mountain sickness are universal at very high altitudes. This review summarizes the pathophysiology of acute mountain sickness and high-altitude headache as well as the evidence for treatment and prevention with different drugs and devices which may be useful for regular and novice mountaineers. This includes an examination of other headache disorders which may mimic high-altitude headache. PMID:25795155

  18. High temperature furnace

    DOEpatents

    Borkowski, Casimer J.

    1976-08-03

    A high temperature furnace for use above 2000.degree.C is provided that features fast initial heating and low power consumption at the operating temperature. The cathode is initially heated by joule heating followed by electron emission heating at the operating temperature. The cathode is designed for routine large temperature excursions without being subjected to high thermal stresses. A further characteristic of the device is the elimination of any ceramic components from the high temperature zone of the furnace.

  19. High temperature refrigerator

    DOEpatents

    Steyert, Jr., William A.

    1978-01-01

    A high temperature magnetic refrigerator which uses a Stirling-like cycle in which rotating magnetic working material is heated in zero field and adiabatically magnetized, cooled in high field, then adiabatically demagnetized. During this cycle said working material is in heat exchange with a pumped fluid which absorbs heat from a low temperature heat source and deposits heat in a high temperature reservoir. The magnetic refrigeration cycle operates at an efficiency 70% of Carnot.

  20. High-Temperature Superconductivity

    NASA Astrophysics Data System (ADS)

    Tanaka, Shoji

    2006-12-01

    A general review on high-temperature superconductivity was made. After prehistoric view and the process of discovery were stated, the special features of high-temperature superconductors were explained from the materials side and the physical properties side. The present status on applications of high-temperature superconductors were explained on superconducting tapes, electric power cables, magnets for maglev trains, electric motors, superconducting quantum interference device (SQUID) and single flux quantum (SFQ) devices and circuits.

  1. High redshift GRBs

    NASA Astrophysics Data System (ADS)

    Gehrels, Neil; Cannizzo, John K.

    2012-09-01

    The Swift mission has opened a new, high redshift window on the universe. In this review we provide an overview of gamma-ray burst (GRB) science, describe the Swift mission, discuss high-z GRBs and tools for high-z studies, and look forward at future capabilities. A new mission concept - Lobster - is described that would monitor the X-ray sky at order of magnitude higher sensitivity than current missions.

  2. High Redshift GRBs

    NASA Technical Reports Server (NTRS)

    Gehrels, Neil; Cannizzo, John K.

    2012-01-01

    The Swift mission has opened a new, high redshift window on the universe. In this review we provide an overview of gamma-ray burst (GRB) science, describe the Swift mission, discuss high-z GRBs and tools for high-z studies, and look forward at future capabilities. A new mission concept - Lobster - is described that would monitor the X-ray sky at order of magnitude higher sensitivity than current missions.

  3. High Reynolds Number Research

    NASA Technical Reports Server (NTRS)

    Baals, D. D. (Editor)

    1977-01-01

    Fundamental aerodynamic questions for which high Reynolds number experimental capability is required are discussed. The operational characteristics and design features of the National Transonic Facility are reviewed.

  4. High altitude decelerator systems

    NASA Technical Reports Server (NTRS)

    Silbert, Mendel N.; Moltedo, A. David; Gilbertson, Gaylord S.

    1989-01-01

    High Altitude Decelerator Systems are used to provide a stable descending platform when deployed from a sounding rocket at altitudes greater than 40 kilometers allowing a scientific mission to be conducted in a specific altitude region. The High Altitude Decelerator is designed to provide a highly stable, high drag area system packed in a minimum volume to deploy successfully from a sounding rocket. Deployment altitudes greater than 100 kilometers have been successfully demonstrated at dynamic pressures as low as 0.004 pounds per square foot.

  5. High power, high frequency helix TWT's

    NASA Astrophysics Data System (ADS)

    Sloley, H. J.; Willard, J.; Paatz, S. R.; Keat, M. J.

    The design and performance characteristics of a 34-GHz pulse tube capable of 75 W peak power output at 30 percent duty cycle and a broadband CW tube are presented. Particular attention is given to the engineering problems encountered during the development of the tubes, including the suppression of backward wave oscillation, the design of electron guns for small-diameter high-current beams, and the thermal capability of small helix structures. The discussion also covers the effects of various design parameters and choice of engineering materials on the ultimate practical limit of power and gain at the operating frequencies. Measurements are presented for advanced experimental tubes.

  6. High power, high frequency, vacuum flange

    DOEpatents

    Felker, B.; McDaniel, M.R.

    1993-03-23

    An improved waveguide flange is disclosed for high power operation that helps prevent arcs from being initiated at the junctions between waveguide sections. The flanges at the end of the waveguide sections have counter bores surrounding the waveguide tubes. When the sections are bolted together the counter bores form a groove that holds a fully annealed copper gasket. Each counterbore has a beveled step that is specially configured to insure the gasket forms a metal-to-metal vacuum seal without gaps or sharp edges. The resultant inner surface of the waveguide is smooth across the junctions between waveguide sections, and arcing is prevented.

  7. High power, high frequency, vacuum flange

    DOEpatents

    Felker, Brian; McDaniel, Michael R.

    1993-01-01

    An improved waveguide flange is disclosed for high power operation that helps prevent arcs from being initiated at the junctions between waveguide sections. The flanges at the end of the waveguide sections have counterbores surrounding the waveguide tubes. When the sections are bolted together the counterbores form a groove that holds a fully annealed copper gasket. Each counterbore has a beveled step that is specially configured to insure the gasket forms a metal-to-metal vacuum seal without gaps or sharp edges. The resultant inner surface of the waveguide is smooth across the junctions between waveguide sections, and arcing is prevented.

  8. High performance, high density hydrocarbon fuels

    NASA Technical Reports Server (NTRS)

    Frankenfeld, J. W.; Hastings, T. W.; Lieberman, M.; Taylor, W. F.

    1978-01-01

    The fuels were selected from 77 original candidates on the basis of estimated merit index and cost effectiveness. The ten candidates consisted of 3 pure compounds, 4 chemical plant streams and 3 refinery streams. Critical physical and chemical properties of the candidate fuels were measured including heat of combustion, density, and viscosity as a function of temperature, freezing points, vapor pressure, boiling point, thermal stability. The best all around candidate was found to be a chemical plant olefin stream rich in dicyclopentadiene. This material has a high merit index and is available at low cost. Possible problem areas were identified as low temperature flow properties and thermal stability. An economic analysis was carried out to determine the production costs of top candidates. The chemical plant and refinery streams were all less than 44 cent/kg while the pure compounds were greater than 44 cent/kg. A literature survey was conducted on the state of the art of advanced hydrocarbon fuel technology as applied to high energy propellents. Several areas for additional research were identified.

  9. Highly Parallel, High-Precision Numerical Integration

    SciTech Connect

    Bailey, David H.; Borwein, Jonathan M.

    2005-04-22

    This paper describes a scheme for rapidly computing numerical values of definite integrals to very high accuracy, ranging from ordinary machine precision to hundreds or thousands of digits, even for functions with singularities or infinite derivatives at endpoints. Such a scheme is of interest not only in computational physics and computational chemistry, but also in experimental mathematics, where high-precision numerical values of definite integrals can be used to numerically discover new identities. This paper discusses techniques for a parallel implementation of this scheme, then presents performance results for 1-D and 2-D test suites. Results are also given for a certain problem from mathematical physics, which features a difficult singularity, confirming a conjecture to 20,000 digit accuracy. The performance rate for this latter calculation on 1024 CPUs is 690 Gflop/s. We believe that this and one other 20,000-digit integral evaluation that we report are the highest-precision non-trivial numerical integrations performed to date.

  10. High Involvement Work Teams.

    ERIC Educational Resources Information Center

    1996

    These three papers were presented at a symposium on high-involvement work teams moderated by Michael Leimbach at the 1996 conference of the Academy of Human Resource Development. "Beyond Training to the New Learning Environment: Workers on the High-Involvement Frontline" (Joseph Anthony Ilacqua, Carol Ann Zulauf) shows the link between an…

  11. High School Chemistry.

    ERIC Educational Resources Information Center

    Journal of Chemical Education, 1983

    1983-01-01

    Preparation for college or life, working conditions and continuing education for high school chemistry teachers, and form/function of high school chemistry textbooks were addressed in presentations at the Seventh Biennial Conference on Chemical Education (Stillwater, Oklahoma 1982). Workshops, lectures, and demonstrations were also presented to…

  12. High Speed data acquisition

    SciTech Connect

    Cooper, P.S.

    1998-02-01

    A general introduction to high Speed data acquisition system techniques in modern particle physics experiments is given. Examples are drawn from the SELEX(E781) high statistics charmed baryon production and decay experiment now taking data at Fermilab. {copyright} {ital 1998 American Institute of Physics.}

  13. High temperature sensor

    DOEpatents

    Tokarz, Richard D.

    1982-01-01

    A high temperature sensor includes a pair of electrical conductors separated by a mass of electrical insulating material. The insulating material has a measurable resistivity within the sensor that changes in relation to the temperature of the insulating material within a high temperature range (1,000 to 2,000 K.). When required, the sensor can be encased within a ceramic protective coating.

  14. High-Stakes Testing.

    ERIC Educational Resources Information Center

    Nieto, Sonia; Irizarry, Jason G.; Luna, Cathy

    2002-01-01

    Discusses four books that provide critical information about the nature of high-stakes tests; the impact of these tests on teaching, learning, and students; the political issues surrounding high-stakes testing; and ways that educators can take informed action to appropriately respond to these tests. (SG)

  15. High School Gifted.

    ERIC Educational Resources Information Center

    Drum, Jean, Ed.

    1993-01-01

    This theme issue on gifted high school students presents several feature articles, a California legislative update, and an editorial. In "Summer Seminar '93--California in the 21st Century," Janis Van Dreal describes a 2-week residential program in which gifted high school students examined California's future. "Visiting a Gifted Program in…

  16. High School Oceanography.

    ERIC Educational Resources Information Center

    Falmouth Public Schools, MA.

    This book is a compilation of a series of papers designed to aid high school teachers in organizing a course in oceanography for high school students. It consists of twelve papers, with references, covering each of the following: (1) Introduction to Oceanography, (2) Geology of the Ocean, (3) The Continental Shelves, (4) Physical Properties of Sea…

  17. High Blood Pressure (Hypertension)

    MedlinePlus

    ... For Consumers Consumer Information by Audience For Women High Blood Pressure (Hypertension) Share Tweet Linkedin Pin it More sharing options ... En Español Who is at risk? How is high blood pressure treated? Understanding your blood pressure: What do the ...

  18. High-Conflict Divorce.

    ERIC Educational Resources Information Center

    Johnston, Janet R.

    1994-01-01

    Reviews available research studies of high-conflict divorce and its effects on children. Factors believed to contribute to high-conflict divorce are explored, and a model of their interrelationships is proposed. Dispute resolution, intervention, and prevention programs are discussed, and implications for social policy are outlined. (SLD)

  19. High energy forming facility

    NASA Technical Reports Server (NTRS)

    Ciurlionis, B.

    1967-01-01

    Watertight, high-explosive forming facility, 25 feet in diameter and 15 feet deep, withstands repeated explosions of 10 pounds of TNT equivalent. The shell is fabricated of high strength steel and allows various structural elements to deform or move elastically and independently while retaining structural integrity.

  20. High School: Erasing Borders

    ERIC Educational Resources Information Center

    Wright, Dianne; Bogotch, Ira

    2006-01-01

    Over the last several years there have been numerous calls for reforming high school to college transitions. In 2000-2001, the American Youth Policy Forum (AYPF) and the National Commission on the High School Senior Year respectively called for a re-thinking of how students moved from secondary to postsecondary education. A widely-discussed…

  1. The American High School

    ERIC Educational Resources Information Center

    Ediger, Marlow

    2008-01-01

    Of all levels of schooling, the high school receives by far the most criticism. There are continuous innovations recommended in journal articles, textbooks, and speeches at state/national conventions on ways to improve the secondary level of schooling. At one teacher education convention, the speaker was criticizing the American high school and…

  2. High School Programming.

    ERIC Educational Resources Information Center

    Update on Gifted Education, 1992

    1992-01-01

    This theme issue covers "High School Programming" and approaches to meeting the needs of gifted high school students in Texas. The first article is "Thematic Interdisciplinary Curricula for Secondary Gifted Students in English and Social Studies" (Joel E. McIntosh and April W. Meacham). This approach is contrasted with the traditional curriculum…

  3. High coking value pitch

    DOEpatents

    Miller, Douglas J.; Chang, Ching-Feng; Lewis, Irwin C.; Lewis, Richard T.

    2014-06-10

    A high coking value pitch prepared from coal tar distillate and has a low softening point and a high carbon value while containing substantially no quinoline insolubles is disclosed. The pitch can be used as an impregnant or binder for producing carbon and graphite articles.

  4. Reforming Underperforming High Schools

    ERIC Educational Resources Information Center

    MDRC, 2013

    2013-01-01

    Urban high schools are in trouble--high dropout rates, low student academic achievement, and graduates who are unprepared for college are just some of the disappointing indicators. However, recent research points to a select number of approaches to improving student outcomes and reforming underperforming schools--from particular ways of creating…

  5. High and Dry

    ERIC Educational Resources Information Center

    Johnson, Robert L.

    2005-01-01

    High-performance schools are facilities that improve the learning environment while saving energy, resources and money. Creating a high-performance school requires an integrated design approach. Key systems--including lighting, HVAC, electrical and plumbing--must be considered from the beginning of the design process. According to William H.…

  6. High current high accuracy IGBT pulse generator

    SciTech Connect

    Nesterov, V.V.; Donaldson, A.R.

    1995-05-01

    A solid state pulse generator capable of delivering high current triangular or trapezoidal pulses into an inductive load has been developed at SLAC. Energy stored in a capacitor bank of the pulse generator is switched to the load through a pair of insulated gate bipolar transistors (IGBT). The circuit can then recover the remaining energy and transfer it back to the capacitor bank without reversing the capacitor voltage. A third IGBT device is employed to control the initial charge to the capacitor bank, a command charging technique, and to compensate for pulse to pulse power losses. The rack mounted pulse generator contains a 525 {mu}F capacitor bank. It can deliver 500 A at 900V into inductive loads up to 3 mH. The current amplitude and discharge time are controlled to 0.02% accuracy by a precision controller through the SLAC central computer system. This pulse generator drives a series pair of extraction dipoles.

  7. High redshift quasars and high metallicities

    NASA Technical Reports Server (NTRS)

    Ferland, Gary J.

    1997-01-01

    A large-scale code called Cloudy was designed to simulate non-equilibrium plasmas and predict their spectra. The goal was to apply it to studies of galactic and extragalactic emission line objects in order to reliably deduce abundances and luminosities. Quasars are of particular interest because they are the most luminous objects in the universe and the highest redshift objects that can be observed spectroscopically, and their emission lines can reveal the composition of the interstellar medium (ISM) of the universe when it was well under a billion years old. The lines are produced by warm (approximately 10(sup 4)K) gas with moderate to low density (n less than or equal to 10(sup 12) cm(sup -3)). Cloudy has been extended to include approximately 10(sup 4) resonance lines from the 495 possible stages of ionization of the lightest 30 elements, an extension that required several steps. The charge transfer database was expanded to complete the needed reactions between hydrogen and the first four ions and fit all reactions with a common approximation. Radiative recombination rate coefficients were derived for recombination from all closed shells, where this process should dominate. Analytical fits to Opacity Project (OP) and other recent photoionization cross sections were produced. Finally, rescaled OP oscillator strengths were used to compile a complete set of data for 5971 resonance lines. The major discovery has been that high redshift quasars have very high metallicities and there is strong evidence that the quasar phenomenon is associated with the birth of massive elliptical galaxies.

  8. High-pressure microfluidics

    NASA Astrophysics Data System (ADS)

    Hjort, K.

    2015-03-01

    When using appropriate materials and microfabrication techniques, with the small dimensions the mechanical stability of microstructured devices allows for processes at high pressures without loss in safety. The largest area of applications has been demonstrated in green chemistry and bioprocesses, where extraction, synthesis and analyses often excel at high densities and high temperatures. This is accessible through high pressures. Capillary chemistry has been used since long but, just like in low-pressure applications, there are several potential advantages in using microfluidic platforms, e.g., planar isothermal set-ups, large local variations in geometries, dense form factors, small dead volumes and precisely positioned microstructures for control of reactions, catalysis, mixing and separation. Other potential applications are in, e.g., microhydraulics, exploration, gas driven vehicles, and high-pressure science. From a review of the state-of-art and frontiers of high pressure microfluidics, the focus will be on different solutions demonstrated for microfluidic handling at high pressures and challenges that remain.

  9. Diamond Ranch High School.

    ERIC Educational Resources Information Center

    Betsky, Aaron

    2000-01-01

    Highlights award-winning Diamond Ranch High School (California) that was designed and built on a steep site around Los Angeles considered unsatisfactory for building due to its unstable soils. Building organization is discussed, and photos are provided. (GR)

  10. High pressure furnace

    DOEpatents

    Morris, D.E.

    1993-09-14

    A high temperature high pressure furnace has a hybrid partially externally heated construction. A metallic vessel fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized (the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum)). The disclosed alloy is fabricated into 11/4 or 2 inch, 32 mm or 50 mm bar stock and has a length of about 22 inches, 56 cm. This bar stock has an aperture formed therein to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the vessel is provided with a small blind aperture into which a thermocouple can be inserted. The closed end of the vessel is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior. 19 figures.

  11. High pressure furnace

    DOEpatents

    Morris, Donald E.

    1993-01-01

    A high temperature high pressure furnace has a hybrid partially externally heated construction. A metallic vessel fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized (the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum). The disclosed alloy is fabricated into 11/4 or 2 inch, 32 mm or 50 mm bar stock and has a length of about 22 inches, 56 cm. This bar stock has an aperture formed therein to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the vessel is provided with a small blind aperture into which a thermocouple can be inserted. The closed end of the vessel is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior.

  12. High blood pressure medicines

    MedlinePlus

    Hypertension - medicines ... blood vessel diseases. You may need to take medicines to lower your blood pressure if lifestyle changes ... blood pressure to the target level. WHEN ARE MEDICINES FOR HIGH BLOOD PRESSURE USED Most of the ...

  13. High-Temperature Lubricants

    NASA Technical Reports Server (NTRS)

    1984-01-01

    In the early 1980's, Lewis Research Center began a program to develop high-temperature lubricants for use on future aircraft flying at three or more times the speed of sound, which can result in vehicle skin temperatures as high as 1,600 degrees Fahrenheit. A material that emerged from this research is a plasma-sprayed, self-lubricating metal- glass-fluoride coating able to reduce oxidation at very high temperatures. Technology is now in commercial use under the trade name Surf-Kote C-800, marketed by Hohman Plating and Manufacturing Inc. and manufactured under a patent license from NASA. Among its uses are lubrication for sliding contact bearings, shaft seals for turbopumps, piston rings for high performance compressors and hot glass processing machinery; it is also widely used in missile and space applications.

  14. High Redshift Quasars

    NASA Technical Reports Server (NTRS)

    Elvis, Martin S.

    1996-01-01

    The report for this period includes three papers: 'Associated Absorption at Low and High Redshift'; 'Strong X-ray Absorption in a Broad Absorption Line Quasar: PHL5200'; and 'ASCA and ROSAT X-ray Spectra of High-Redshift Radio-Loud Quasars'. The first gives examples from both low and high redshift for combining information on absorbing material in active galactic nuclei from both x-ray and the UV. The second presents ASCA observations of the z = 1.98 prototype broad absorption line quasar (BALQSO): PHL 5200, detected with both the solid-state imaging spectrometers and the gas imaging spectometers. The third paper presents results on the x-ray properties of 9 high-redshift radio-loud quasars observed by ASCA and ROSAT, including ASCA observations of S5 0014+81 (z = 3.38) and S5 0836+71 (z = 2.17) and ROSAT observations of PKS 2126-158.

  15. High-Temperature Superconductivity

    SciTech Connect

    Peter Johnson

    2008-11-05

    Like astronomers tweaking images to gain a more detailed glimpse of distant stars, physicists at Brookhaven National Laboratory have found ways to sharpen images of the energy spectra in high-temperature superconductors — materials that carry electrical c

  16. Highly Autonomous Systems Workshop

    NASA Technical Reports Server (NTRS)

    Doyle, R.; Rasmussen, R.; Man, G.; Patel, K.

    1998-01-01

    It is our aim by launching a series of workshops on the topic of highly autonomous systems to reach out to the larger community interested in technology development for remotely deployed systems, particularly those for exploration.

  17. High Power Hall Thrusters

    NASA Technical Reports Server (NTRS)

    Jankovsky, Robert; Tverdokhlebov, Sergery; Manzella, David

    1999-01-01

    The development of Hall thrusters with powers ranging from tens of kilowatts to in excess of one hundred kilowatts is considered based on renewed interest in high power. high thrust electric propulsion applications. An approach to develop such thrusters based on previous experience is discussed. It is shown that the previous experimental data taken with thrusters of 10 kW input power and less can be used. Potential mass savings due to the design of high power Hall thrusters are discussed. Both xenon and alternate thruster propellant are considered, as are technological issues that will challenge the design of high power Hall thrusters. Finally, the implications of such a development effort with regard to ground testing and spacecraft intecrati'on issues are discussed.

  18. High-mix insulins

    PubMed Central

    Kalra, Sanjay; Farooqi, Mohammad Hamed; El-Houni, Ali E.

    2015-01-01

    Premix insulins are commonly used insulin preparations, which are available in varying ratios of different molecules. These drugs contain one short- or rapid-acting, and one intermediate- or long-acting insulin. High-mix insulins are mixtures of insulins that contain 50% or more than 50% of short-acting insulin. This review describes the clinical pharmacology of high-mix insulins, including data from randomized controlled trials. It suggests various ways, in which high-mix insulin can be used, including once daily, twice daily, thrice daily, hetero-mix, and reverse regimes. The authors provide a rational framework to help diabetes care professionals, identify indications for pragmatic high-mix use. PMID:26425485

  19. Hypertension (High Blood Pressure)

    MedlinePlus

    ... pressure to live. Without it, blood can't flow through our bodies and carry oxygen to our vital organs. But when blood pressure gets too high — a condition called hypertension — it can lead to ...

  20. High Blood Pressure

    MedlinePlus

    ... Division of Geriatrics and Clinical Gerontology Division of Neuroscience FAQs Funding Opportunities Intramural Research Program Office of ... to major health problems. Make a point of learning what blood pressure should be. And, remember: High ...

  1. High blood cholesterol levels

    MedlinePlus

    ... gov/ency/article/000403.htm High blood cholesterol levels To use the sharing features on this page, ... called "bad" cholesterol For many people, abnormal cholesterol levels are partly due to an unhealthy lifestyle. This ...

  2. High pressure oxygen furnace

    DOEpatents

    Morris, Donald E.

    1992-01-01

    A high temperature high pressure oxygen furnace having a hybrid partially externally heated construction is disclosed. A metallic bar fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized (the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum). The disclosed alloy is fabricated into 11/4 inch bar stock and has a length of about 17 inches. This bar stock is gun drilled for over 16 inches of its length with 0.400 inch aperture to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the bar is provided with a small support aperture into which both a support and a thermocouple can be inserted. The closed end of the gun drilled bar is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior.

  3. High pressure oxygen furnace

    DOEpatents

    Morris, D.E.

    1992-07-14

    A high temperature high pressure oxygen furnace having a hybrid partially externally heated construction is disclosed. A metallic bar fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized, the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum). The disclosed alloy is fabricated into 11/4 inch bar stock and has a length of about 17 inches. This bar stock is gun drilled for over 16 inches of its length with 0.400 inch aperture to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the bar is provided with a small support aperture into which both a support and a thermocouple can be inserted. The closed end of the gun drilled bar is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior. 5 figs.

  4. High Velocity Gas Gun

    NASA Technical Reports Server (NTRS)

    1988-01-01

    A video tape related to orbital debris research is presented. The video tape covers the process of loading a High Velocity Gas Gun and firing it into a mounted metal plate. The process is then repeated in slow motion.

  5. High-Temperature Superconductivity

    ScienceCinema

    Peter Johnson

    2010-01-08

    Like astronomers tweaking images to gain a more detailed glimpse of distant stars, physicists at Brookhaven National Laboratory have found ways to sharpen images of the energy spectra in high-temperature superconductors ? materials that carry electrical c

  6. Hyperglycemia (High Blood Glucose)

    MedlinePlus Videos and Cool Tools

    ... and Learning About Prediabetes Type 2 Diabetes Risk Test Lower Your Risk Healthy Eating Overweight Smoking High ... prediabetes, and no one is excused. Take the test. Know where you stand. sticky en -- Chef Ronaldo's ...

  7. High School Parking Lots.

    ERIC Educational Resources Information Center

    Neff, Thomas G.

    2002-01-01

    Describes the reorganization of the site of Ben Davis High School in Wayne Township, Indiana as an example of improvements to school parking lot design and vehicle/pedestrian traffic flow and security. Includes design drawings. (EV)

  8. High Speed Research Program

    NASA Technical Reports Server (NTRS)

    Anderson, Robert E.; Corsiglia, Victor R.; Schmitz, Frederic H. (Technical Monitor)

    1994-01-01

    An overview of the NASA High Speed Research Program will be presented from a NASA Headquarters perspective. The presentation will include the objectives of the program and an outline of major programmatic issues.

  9. High Efficiency, High Performance Clothes Dryer

    SciTech Connect

    Peter Pescatore; Phil Carbone

    2005-03-31

    This program covered the development of two separate products; an electric heat pump clothes dryer and a modulating gas dryer. These development efforts were independent of one another and are presented in this report in two separate volumes. Volume 1 details the Heat Pump Dryer Development while Volume 2 details the Modulating Gas Dryer Development. In both product development efforts, the intent was to develop high efficiency, high performance designs that would be attractive to US consumers. Working with Whirlpool Corporation as our commercial partner, TIAX applied this approach of satisfying consumer needs throughout the Product Development Process for both dryer designs. Heat pump clothes dryers have been in existence for years, especially in Europe, but have not been able to penetrate the market. This has been especially true in the US market where no volume production heat pump dryers are available. The issue has typically been around two key areas: cost and performance. Cost is a given in that a heat pump clothes dryer has numerous additional components associated with it. While heat pump dryers have been able to achieve significant energy savings compared to standard electric resistance dryers (over 50% in some cases), designs to date have been hampered by excessively long dry times, a major market driver in the US. The development work done on the heat pump dryer over the course of this program led to a demonstration dryer that delivered the following performance characteristics: (1) 40-50% energy savings on large loads with 35 F lower fabric temperatures and similar dry times; (2) 10-30 F reduction in fabric temperature for delicate loads with up to 50% energy savings and 30-40% time savings; (3) Improved fabric temperature uniformity; and (4) Robust performance across a range of vent restrictions. For the gas dryer development, the concept developed was one of modulating the gas flow to the dryer throughout the dry cycle. Through heat modulation in a

  10. High efficiency incandescent lighting

    SciTech Connect

    Bermel, Peter; Ilic, Ognjen; Chan, Walker R.; Musabeyoglu, Ahmet; Cukierman, Aviv Ruben; Harradon, Michael Robert; Celanovic, Ivan; Soljacic, Marin

    2014-09-02

    Incandescent lighting structure. The structure includes a thermal emitter that can, but does not have to, include a first photonic crystal on its surface to tailor thermal emission coupled to, in a high-view-factor geometry, a second photonic filter selected to reflect infrared radiation back to the emitter while passing visible light. This structure is highly efficient as compared to standard incandescent light bulbs.

  11. High conductivity composite metal

    DOEpatents

    Zhou, Ruoyi; Smith, James L.; Embury, John David

    1998-01-01

    Electrical conductors and methods of producing them, where the conductors possess both high strength and high conductivity. Conductors are comprised of carbon steel and a material chosen from a group consisting of copper, nickel, silver, and gold. Diffusion barriers are placed between these two materials. The components of a conductor are assembled and then the assembly is subjected to heat treating and mechanical deformation steps.

  12. High conductivity composite metal

    DOEpatents

    Zhou, R.; Smith, J.L.; Embury, J.D.

    1998-01-06

    Electrical conductors and methods of producing them are disclosed, where the conductors possess both high strength and high conductivity. Conductors are comprised of carbon steel and a material chosen from a group consisting of copper, nickel, silver, and gold. Diffusion barriers are placed between these two materials. The components of a conductor are assembled and then the assembly is subjected to heat treating and mechanical deformation steps. 10 figs.

  13. High pressure gas target

    NASA Astrophysics Data System (ADS)

    Gelbart, W.; Johnson, R. R.; Abeysekera, B.

    2012-12-01

    Compact, high pressure, high current gas target features all metal construction and semi-automatic window assembly change. The unique aspect of this target is the domed-shaped window. The Havar alloy window is electron beam welded to a metal ring, thus forming one, interchangeable assembly. The window assembly is sealed by knife-edges locked by a pneumatic toggle allowing a quick, in situ window change.

  14. High beta multipoles

    SciTech Connect

    Prager, S C

    1982-05-01

    Multipoles are being employed as devices to study fusion issues and plasma phenomena at high values of beta (plasma pressure/magnetic pressure) in a controlled manner. Due to their large volume, low magnetic field (low synchrotron radiation) region, they are also under consideration as potential steady state advanced fuel (low neutron yield) reactors. Present experiments are investigating neoclassical (bootstrap and Pfirsch-Schlueter) currents and plasma stability at extremely high beta.

  15. High temperature pressure gauge

    DOEpatents

    Echtler, J. Paul; Scandrol, Roy O.

    1981-01-01

    A high temperature pressure gauge comprising a pressure gauge positioned in fluid communication with one end of a conduit which has a diaphragm mounted in its other end. The conduit is filled with a low melting metal alloy above the diaphragm for a portion of its length with a high temperature fluid being positioned in the remaining length of the conduit and in the pressure gauge.

  16. High Performance Polymers

    NASA Technical Reports Server (NTRS)

    Venumbaka, Sreenivasulu R.; Cassidy, Patrick E.

    2003-01-01

    This report summarizes results from research on high performance polymers. The research areas proposed in this report include: 1) Effort to improve the synthesis and to understand and replicate the dielectric behavior of 6HC17-PEK; 2) Continue preparation and evaluation of flexible, low dielectric silicon- and fluorine- containing polymers with improved toughness; and 3) Synthesis and characterization of high performance polymers containing the spirodilactam moiety.

  17. High-temperature electronics

    NASA Technical Reports Server (NTRS)

    Matus, Lawrence G.; Seng, Gary T.

    1990-01-01

    To meet the needs of the aerospace propulsion and space power communities, the high temperature electronics program at the Lewis Research Center is developing silicon carbide (SiC) as a high temperature semiconductor material. This program supports a major element of the Center's mission - to perform basic and developmental research aimed at improving aerospace propulsion systems. Research is focused on developing the crystal growth, characterization, and device fabrication technologies necessary to produce a family of SiC devices.

  18. HIGH TEMPERATURE THERMOCOUPLE

    DOEpatents

    Eshayu, A.M.

    1963-02-12

    This invention contemplates a high temperature thermocouple for use in an inert or a reducing atmosphere. The thermocouple limbs are made of rhenium and graphite and these limbs are connected at their hot ends in compressed removable contact. The rhenium and graphite are of high purity and are substantially stable and free from diffusion into each other even without shielding. Also, the graphite may be thick enough to support the thermocouple in a gas stream. (AEC)

  19. High-temperature sensor

    DOEpatents

    Not Available

    1981-01-29

    A high temperature sensor is described which includes a pair of electrical conductors separated by a mass of electrical insulating material. The insulating material has a measurable resistivity within the sensor that changes in relation to the temperature of the insulating material within a high temperature range (1000 to 2000/sup 0/K). When required, the sensor can be encased within a ceramic protective coating.

  20. High-temperature electronics

    NASA Astrophysics Data System (ADS)

    Matus, Lawrence G.; Seng, Gary T.

    1990-02-01

    To meet the needs of the aerospace propulsion and space power communities, the high temperature electronics program at the Lewis Research Center is developing silicon carbide (SiC) as a high temperature semiconductor material. This program supports a major element of the Center's mission - to perform basic and developmental research aimed at improving aerospace propulsion systems. Research is focused on developing the crystal growth, characterization, and device fabrication technologies necessary to produce a family of SiC devices.

  1. High performance polymeric foams

    SciTech Connect

    Gargiulo, M.; Sorrentino, L.; Iannace, S.

    2008-08-28

    The aim of this work was to investigate the foamability of high-performance polymers (polyethersulfone, polyphenylsulfone, polyetherimide and polyethylenenaphtalate). Two different methods have been used to prepare the foam samples: high temperature expansion and two-stage batch process. The effects of processing parameters (saturation time and pressure, foaming temperature) on the densities and microcellular structures of these foams were analyzed by using scanning electron microscopy.

  2. Fulminant high altitude blindness.

    PubMed

    Mashkovskiy, Evgeny; Szawarski, Piotr; Ryzhkov, Pavel; Goslar, Tomaz; Mrak, Irena

    2016-06-01

    Prolonged altitude exposure even with acclimatization continues to present a physiological challenge to all organ systems including the central nervous system. We describe a case of a 41-year-old Caucasian female climber who suffered severe visual loss that was due to possible optic nerve pathology occurring during a high altitude expedition in the Himalayas. This case is atypical of classic high altitude cerebral oedema and highlights yet another danger of prolonged sojourn at extreme altitudes. PMID:27601532

  3. High power density targets

    NASA Astrophysics Data System (ADS)

    Pellemoine, Frederique

    2013-12-01

    In the context of new generation rare isotope beam facilities based on high-power heavy-ion accelerators and in-flight separation of the reaction products, the design of the rare isotope production targets is a major challenge. In order to provide high-purity beams for science, high resolution is required in the rare isotope separation. This demands a small beam spot on the production target which, together with the short range of heavy ions in matter, leads to very high power densities inside the target material. This paper gives an overview of the challenges associated with this high power density, discusses radiation damage issues in targets exposed to heavy ion beams, and presents recent developments to meet some of these challenges through different projects: FAIR, RIBF and FRIB which is the most challenging. Extensive use of Finite Element Analysis (FEA) has been made at all facilities to specify critical target parameters and R&D work at FRIB successfully retired two major risks related to high-power density and heavy-ion induced radiation damage.

  4. High Energy Continuum of High Redshift Quasars

    NASA Technical Reports Server (NTRS)

    Elvis, Martin

    2000-01-01

    Discussion with the RXTE team at GSFC showed that a sufficiently accurate background subtraction procedure had now, been derived for sources at the flux level of PKS 2126-158. However this solution does not apply to observations carried out before April 1997, including our observation. The prospect of an improved solution becoming available soon is slim. As a result the RXTE team agreed to re-observe PKS2126-158. The new observation was carried out in April 1999. Quasi-simultaneous optical observations were obtained, as Service observing., at the 4-meter Anglo-Australian Telescope, and ftp-ed from the AAT on 22April. The RXTE data was processed in late June, arriving at SAO in early July. Coincidentally, our collaborative Beppo-SAX observation of PKS2126-158 was made later in 1999, and a GTO Chandra observation (with which we are involved) was made on November 16. Since this gives us a unique monitoring data for a high redshift quasar over a broad pass-band we are now combining all three observations into a single comprehensive study Final publication of the RXTE data will thus take place under another grant.

  5. High Performance Arcjet Engines

    NASA Technical Reports Server (NTRS)

    Kennel, Elliot B.; Ivanov, Alexey Nikolayevich; Nikolayev, Yuri Vyacheslavovich

    1994-01-01

    This effort sought to exploit advanced single crystal tungsten-tantalum alloy material for fabrication of a high strength, high temperature arcjet anode. The use of this material is expected to result in improved strength, temperature resistance, and lifetime compared to state of the art polycrystalline alloys. In addition, the use of high electrical and thermal conductivity carbon-carbon composites was considered, and is believed to be a feasible approach. Highly conductive carbon-carbon composite anode capability represents enabling technology for rotating-arc designs derived from the Russian Scientific Research Institute of Thermal Processes (NIITP) because of high heat fluxes at the anode surface. However, for US designs the anode heat flux is much smaller, and thus the benefits are not as great as in the case of NIITP-derived designs. Still, it does appear that the tensile properties of carbon-carbon can be even better than those of single crystal tungsten alloys, especially when nearly-single-crystal fibers such as vapor grown carbon fiber (VGCF) are used. Composites fabricated from such materials must be coated with a refractory carbide coating in order to ensure compatibility with high temperature hydrogen. Fabrication of tungsten alloy single crystals in the sizes required for fabrication of an arcjet anode has been shown to be feasible. Test data indicate that the material can be expected to be at least the equal of W-Re-HfC polycrystalline alloy in terms of its tensile properties, and possibly superior. We are also informed by our colleagues at Scientific Production Association Luch (NP0 Luch) that it is possible to use Russian technology to fabricate polycrystalline W-Re-HfC or other high strength alloys if desired. This is important because existing engines must rely on previously accumulated stocks of these materials, and a fabrication capability for future requirements is not assured.

  6. High Beta Tokamaks

    SciTech Connect

    Cowley, S.

    1998-11-14

    Perhaps the ideal tokamak would have high {beta} ({beta} {approx}> 1) and classical confinement. Such a tokamak has not been found, and we do not know if one does exist. We have searched for such a possibility, so far without success. In 1990, we obtained analytic equilibrium solutions for large aspect ratio tokamaks at {beta} {approx} {Omicron}(1) [1]. These solutions and the extension at high {beta} poloidal to finite aspect ratio [2] provided a basis for the study of high {beta} tokamaks. We have shown that these configurations can be stable to short scale MHD modes [3], and that they have reduced neoclassical transport [4]. Microinstabilities (such as the {del}T{sub i} mode) seem to be stabilized at high {beta} [5] - this is due to the large local shear [3] and the magnetic well. We have some concerns about modes associated with the compressional branch which may appear at high {beta}. Bill Dorland and Mike Kotschenreuther have studied this issue and our concerns may be unfounded. It is certainly tantalizing, especially given the lowered neoclassical transport values, that these configurations could have no microinstabilities and, one could assume, no anomalous transport. Unfortunately, while this work is encouraging, the key question for high {beta} tokamaks is the stability to large scale kink modes. The MHD {beta} limit (Troyon limit) for kink modes at large aspect ratio is problematically low. There is ample evidence from computations that the limit exists. However, it is not known if stable equilibria exist at much higher {beta}--none have been found. We have explored this question in the asymptotic high {beta} poloidal limit. Unfortunately, we are unable to find stable equilibrium and also unable to show that they don't exist. The results of these calculations will be published when a more definitive answer is found.

  7. Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

    SciTech Connect

    Ingersoll, DT

    2005-12-15

    A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of forced circulation

  8. The effect of fission products on the rate of U3O8 formation in SIMFUEL oxidized in air at 250°C

    NASA Astrophysics Data System (ADS)

    Choi, Jong-Won; McEachern, Rod J.; Taylor, Peter; Wood, Donald D.

    1996-06-01

    The effect of fission products on the rate of U3O8 formation was investigated by oxidizing UO2-based SIMFUEL (simulated high burnup nuclear fuel) and unirradiated UO2 fuel specimens in air at 250°C for different times (1-317 days). The progress of oxidation was monitored by X-ray diffraction, revealing that the rate of U3O8 formation declines with increasing burnup. An expression was derived to describe quantitatively the time for U3O8 powder formation as a function of simulated burnup. These findings were supported by additional isochronal oxidation experiments conducted between 200 and 300°C.

  9. PRODUCTION OF PLUTONIUM METAL

    DOEpatents

    Lyon, W.L.; Moore, R.H.

    1961-01-17

    A process is given for producing plutonium metal by the reduction of plutonium chloride, dissolved in alkali metal chloride plus or minus aluminum chloride, with magnesium or a magnesium-aluminum alloy at between 700 and 800 deg C and separating the plutonium or plutonium-aluminum alloy formed from the salt.

  10. High Temperature Capacitor Development

    SciTech Connect

    John Kosek

    2009-06-30

    The absence of high-temperature electronics is an obstacle to the development of untapped energy resources (deep oil, gas and geothermal). US natural gas consumption is projected to grow from 22 trillion cubic feet per year (tcf) in 1999 to 34 tcf in 2020. Cumulatively this is 607 tcf of consumption by 2020, while recoverable reserves using current technology are 177 tcf. A significant portion of this shortfall may be met by tapping deep gas reservoirs. Tapping these reservoirs represents a significant technical challenge. At these depths, temperatures and pressures are very high and may require penetrating very hard rock. Logistics of supporting 6.1 km (20,000 ft) drill strings and the drilling processes are complex and expensive. At these depths up to 50% of the total drilling cost may be in the last 10% of the well depth. Thus, as wells go deeper it is increasingly important that drillers are able to monitor conditions down-hole such as temperature, pressure, heading, etc. Commercial off-the-shelf electronics are not specified to meet these operating conditions. This is due to problems associated with all aspects of the electronics including the resistors and capacitors. With respect to capacitors, increasing temperature often significantly changes capacitance because of the strong temperature dependence of the dielectric constant. Higher temperatures also affect the equivalent series resistance (ESR). High-temperature capacitors usually have low capacitance values because of these dielectric effects and because packages are kept small to prevent mechanical breakage caused by thermal stresses. Electrolytic capacitors do not operate at temperatures above 150oC due to dielectric breakdown. The development of high-temperature capacitors to be used in a high-pressure high-temperature (HPHT) drilling environment was investigated. These capacitors were based on a previously developed high-voltage hybridized capacitor developed at Giner, Inc. in conjunction with a

  11. High Altitude Medical Problems

    PubMed Central

    Hultgren, Herbert N.

    1979-01-01

    Increased travel to high altitude areas by mountaineers and nonclimbing tourists has emphasized the clinical problems associated with rapid ascent. Acute mountain sickness affects most sojourners at elevations above 10,000 feet. Symptoms are usually worse on the second or third day after arrival. Gradual ascent, spending one to three days at an intermediate altitude, and the use of acetazolamide (Diamox) will prevent or ameliorate symptoms in most instances. Serious and potentially fatal problems, such as high altitude pulmonary edema or cerebral edema, occur in approximately 0.5 percent to 1.0 percent of visitors to elevations above 10,000 feet—especially with heavy physical exertion on arrival, such as climbing or skiing. Early recognition, high flow oxygen therapy and prompt descent are crucially important in management. Our knowledge of the causes of these and other high altitude problems, such as retinal hemorrhage, systemic edema and pulmonary hypertension, is still incomplete. Even less is known of the effect of high altitudes on medical conditions common at sea level or on the action of commonly used drugs. ImagesFigure 2. PMID:483805

  12. High level nuclear waste

    SciTech Connect

    Crandall, J L

    1980-01-01

    The DOE Division of Waste Products through a lead office at Savannah River is developing a program to immobilize all US high-level nuclear waste for terminal disposal. DOE high-level wastes include those at the Hanford Plant, the Idaho Chemical Processing Plant, and the Savannah River Plant. Commercial high-level wastes, for which DOE is also developing immobilization technology, include those at the Nuclear Fuel Services Plant and any future commercial fuels reprocessing plants. The first immobilization plant is to be the Defense Waste Processing Facility at Savannah River, scheduled for 1983 project submission to Congress and 1989 operation. Waste forms are still being selected for this plant. Borosilicate glass is currently the reference form, but alternate candidates include concretes, calcines, other glasses, ceramics, and matrix forms.

  13. High brightness electron accelerator

    DOEpatents

    Sheffield, Richard L.; Carlsten, Bruce E.; Young, Lloyd M.

    1994-01-01

    A compact high brightness linear accelerator is provided for use, e.g., in a free electron laser. The accelerator has a first plurality of acclerating cavities having end walls with four coupling slots for accelerating electrons to high velocities in the absence of quadrupole fields. A second plurality of cavities receives the high velocity electrons for further acceleration, where each of the second cavities has end walls with two coupling slots for acceleration in the absence of dipole fields. The accelerator also includes a first cavity with an extended length to provide for phase matching the electron beam along the accelerating cavities. A solenoid is provided about the photocathode that emits the electons, where the solenoid is configured to provide a substantially uniform magnetic field over the photocathode surface to minimize emittance of the electons as the electrons enter the first cavity.

  14. HIGH POWER PULSED OSCILLATOR

    DOEpatents

    Singer, S.; Neher, L.K.

    1957-09-24

    A high powered, radio frequency pulse oscillator is described for generating trains of oscillations at the instant an input direct voltage is impressed, or immediately upon application of a light pulse. In one embodiment, the pulse oscillator comprises a photo-multiplier tube with the cathode connected to the first dynode by means of a resistor, and adjacent dynodes are connected to each other through adjustable resistors. The ohmage of the resistors progressively increases from a very low value for resistors adjacent the cathode to a high value adjacent the plate, the last dynode. Oscillation occurs with this circuit when a high negative voltage pulse is applied to the cathode and the photo cathode is bombarded. Another embodiment adds capacitors at the resistor connection points of the above circuit to increase the duration of the oscillator train.

  15. Bioremediation of high explosives

    SciTech Connect

    Kitts, C.L.; Alvarez, M.A.; Hanners, J.L.; Ogden, K.L.; Vanderberg-Twary, L.; Unkefer, P.J.

    1995-09-01

    Manufacture and use of high explosives has resulted in contamination of ground water and soils throughout the world. The use of biological methods for remediation of high explosives contamination has received considerable attention in recent years. Biodegradation is most easily studied using organisms in liquid cultures. Thus, the amount of explosive that can be degraded in liquid culture is quite small. However, these experiments are useful for gathering basic information about the biochemical pathways of biodegradation, identifying appropriate organisms and obtaining rates of degradation. The authors` laboratory has investigated all three major areas of explosives bioremediation: explosives in solution, explosives in soil, and the disposal of bulk explosives from demilitarization operations. They investigated the three explosives most commonly used in modern high explosive formulations: 2,4,6-trinitrotoluene (TNT), hexahydro 1,3,5-trinitro-1,3,5-triazine (RDX) and octahydro-1,3,5,7-tetranitro-1,3,5,7-tetrazocine (HMX).

  16. Highly Thermal Conductive Nanocomposites

    NASA Technical Reports Server (NTRS)

    Sun, Ya-Ping (Inventor); Connell, John W. (Inventor); Veca, Lucia Monica (Inventor)

    2015-01-01

    Disclosed are methods for forming carbon-based fillers as may be utilized in forming highly thermal conductive nanocomposite materials. Formation methods include treatment of an expanded graphite with an alcohol/water mixture followed by further exfoliation of the graphite to form extremely thin carbon nanosheets that are on the order of between about 2 and about 10 nanometers in thickness. Disclosed carbon nanosheets can be functionalized and/or can be incorporated in nanocomposites with extremely high thermal conductivities. Disclosed methods and materials can prove highly valuable in many technological applications including, for instance, in formation of heat management materials for protective clothing and as may be useful in space exploration or in others that require efficient yet light-weight and flexible thermal management solutions.

  17. High Performance, Dependable Multiprocessor

    NASA Technical Reports Server (NTRS)

    Ramos, Jeremy; Samson, John R.; Troxel, Ian; Subramaniyan, Rajagopal; Jacobs, Adam; Greco, James; Cieslewski, Grzegorz; Curreri, John; Fischer, Michael; Grobelny, Eric; George, Alan; Aggarwal, Vikas; Patel, Minesh; Some, Raphael

    2006-01-01

    With the ever increasing demand for higher bandwidth and processing capacity of today's space exploration, space science, and defense missions, the ability to efficiently apply commercial-off-the-shelf (COTS) processors for on-board computing is now a critical need. In response to this need, NASA's New Millennium Program office has commissioned the development of Dependable Multiprocessor (DM) technology for use in payload and robotic missions. The Dependable Multiprocessor technology is a COTS-based, power efficient, high performance, highly dependable, fault tolerant cluster computer. To date, Honeywell has successfully demonstrated a TRL4 prototype of the Dependable Multiprocessor [I], and is now working on the development of a TRLS prototype. For the present effort Honeywell has teamed up with the University of Florida's High-performance Computing and Simulation (HCS) Lab, and together the team has demonstrated major elements of the Dependable Multiprocessor TRLS system.

  18. High-conflict divorce.

    PubMed

    Johnston, J R

    1994-01-01

    This article reviews available research studies of high-conflict divorce and its effects on children. Interparental conflict after divorce (defined as verbal and physical aggression, overt hostility, and distrust) and the primary parent's emotional distress are jointly predictive of more problematic parent-child relationships and greater child emotional and behavioral maladjustment. As a group, children of high-conflict divorce as defined above, especially boys, are two to four times more likely to be clinically disturbed in emotions and behavior compared with national norms. Court-ordered joint physical custody and frequent visitation arrangements in high-conflict divorce tend to be associated with poorer child outcomes, especially for girls. Types of intervention programs and social policy appropriate for these kinds of families are presented. PMID:7922278

  19. High temperature interface superconductivity

    DOE PAGESBeta

    Gozar, A.; Bozovic, I.

    2016-01-20

    High-Tc superconductivity at interfaces has a history of more than a couple of decades. In this review we focus our attention on copper-oxide based heterostructures and multi-layers. We first discuss the technique, atomic layer-by-layer molecular beam epitaxy (ALL-MBE) engineering, that enabled High-Tc Interface Superconductivity (HT-IS), and the challenges associated with the realization of high quality interfaces. Then we turn our attention to the experiments which shed light on the structure and properties of interfacial layers, allowing comparison to those of single-phase films and bulk crystals. Both ‘passive’ hetero-structures as well as surface-induced effects by external gating are discussed. Here, wemore » conclude by comparing HT-IS in cuprates and in other classes of materials, especially Fe-based superconductors, and by examining the grand challenges currently laying ahead for the field.« less

  20. High voltage power supply

    NASA Technical Reports Server (NTRS)

    Ruitberg, A. P.; Young, K. M. (Inventor)

    1985-01-01

    A high voltage power supply is formed by three discrete circuits energized by a battery to provide a plurality of concurrent output signals floating at a high output voltage on the order of several tens of kilovolts. In the first two circuits, the regulator stages are pulse width modulated and include adjustable ressistances for varying the duty cycles of pulse trains provided to corresponding oscillator stages while the third regulator stage includes an adjustable resistance for varying the amplitude of a steady signal provided to a third oscillator stage. In the first circuit, the oscillator, formed by a constant current drive network and a tuned resonant network included a step up transformer, is coupled to a second step up transformer which, in turn, supplies an amplified sinusoidal signal to a parallel pair of complementary poled rectifying, voltage multiplier stages to generate the high output voltage.

  1. High resolution data acquisition

    DOEpatents

    Thornton, Glenn W.; Fuller, Kenneth R.

    1993-01-01

    A high resolution event interval timing system measures short time intervals such as occur in high energy physics or laser ranging. Timing is provided from a clock (38) pulse train (37) and analog circuitry (44) for generating a triangular wave (46) synchronously with the pulse train (37). The triangular wave (46) has an amplitude and slope functionally related to the time elapsed during each clock pulse in the train. A converter (18, 32) forms a first digital value of the amplitude and slope of the triangle wave at the start of the event interval and a second digital value of the amplitude and slope of the triangle wave at the end of the event interval. A counter (26) counts the clock pulse train (37) during the interval to form a gross event interval time. A computer (52) then combines the gross event interval time and the first and second digital values to output a high resolution value for the event interval.

  2. High resolution data acquisition

    DOEpatents

    Thornton, G.W.; Fuller, K.R.

    1993-04-06

    A high resolution event interval timing system measures short time intervals such as occur in high energy physics or laser ranging. Timing is provided from a clock, pulse train, and analog circuitry for generating a triangular wave synchronously with the pulse train (as seen in diagram on patent). The triangular wave has an amplitude and slope functionally related to the time elapsed during each clock pulse in the train. A converter forms a first digital value of the amplitude and slope of the triangle wave at the start of the event interval and a second digital value of the amplitude and slope of the triangle wave at the end of the event interval. A counter counts the clock pulse train during the interval to form a gross event interval time. A computer then combines the gross event interval time and the first and second digital values to output a high resolution value for the event interval.

  3. High temperature electronics technology

    NASA Astrophysics Data System (ADS)

    Dening, J. C.; Hurtle, D. E.

    1984-03-01

    This report summarizes the barrier metallization developments accomplished in a program intended to develop 300 C electronic controls capability for potential on-engine aircraft engine application. In addition, this report documents preliminary life test results at 300 C and above and discusses improved design practices required for high temperature integrated injection logic semiconductors. Previous Phase 1 activities focused on determining the viability of operating silicon semiconductor devices over the -55 C to +300 C temperature range. This feasibility was substantiated but the need for additional design work and process development was indicated. Phase 2 emphasized the development of a high temperature metallization system as the primary development need for high temperature silicon semiconductor applications.

  4. High temperature interface superconductivity

    NASA Astrophysics Data System (ADS)

    Gozar, A.; Bozovic, I.

    2016-02-01

    High-Tc superconductivity at interfaces has a history of more than a couple of decades. In this review we focus our attention on copper-oxide based heterostructures and multi-layers. We first discuss the technique, atomic layer-by-layer molecular beam epitaxy (ALL-MBE) engineering, that enabled High-Tc Interface Superconductivity (HT-IS), and the challenges associated with the realization of high quality interfaces. Then we turn our attention to the experiments which shed light on the structure and properties of interfacial layers, allowing comparison to those of single-phase films and bulk crystals. Both 'passive' hetero-structures as well as surface-induced effects by external gating are discussed. We conclude by comparing HT-IS in cuprates and in other classes of materials, especially Fe-based superconductors, and by examining the grand challenges currently laying ahead for the field.

  5. High-energy detector

    DOEpatents

    Bolotnikov, Aleksey E.; Camarda, Giuseppe; Cui, Yonggang; James, Ralph B.

    2011-11-22

    The preferred embodiments are directed to a high-energy detector that is electrically shielded using an anode, a cathode, and a conducting shield to substantially reduce or eliminate electrically unshielded area. The anode and the cathode are disposed at opposite ends of the detector and the conducting shield substantially surrounds at least a portion of the longitudinal surface of the detector. The conducting shield extends longitudinally to the anode end of the detector and substantially surrounds at least a portion of the detector. Signals read from one or more of the anode, cathode, and conducting shield can be used to determine the number of electrons that are liberated as a result of high-energy particles impinge on the detector. A correction technique can be implemented to correct for liberated electron that become trapped to improve the energy resolution of the high-energy detectors disclosed herein.

  6. High School Principals and the High School Journalism Program.

    ERIC Educational Resources Information Center

    Peterson, Jane W.

    A study asked selected high school principals to respond to statements about the value of high school journalism to the high school student and about the rights and responsibilities of the high school journalist. These responses were then checked against such information as whether or not the high school principal had worked on a high school…

  7. High temperature electronics

    NASA Astrophysics Data System (ADS)

    Seng, Gary T.

    1991-03-01

    In recent years, the aerospace propulsion and space power communities have acknowledged a growing need for electronic devices that are capable of sustained high-temperature operation. Aeropropulsion applications for high-temperature electronic devices include engine ground test instrumentation such as multiplexers, analog-to-digital converters, and telemetry systems capable of withstanding hot section engine temperatures in excess of 600 C. Uncooled operation of control and condition monitoring systems in advanced supersonic aircraft would subject the electronics to temperatures in excess of 300 C. Similarly, engine-mounted integrated electronic sensors could reach temperatures which exceed 500 C. In addition to aeronautics, there are many other areas that could benefit from the existence of high-temperature electronic devices. Space applications include power electronic devices for space platforms and satellites. Since power electronics require radiators to shed waste heat, electronic devices that operate at higher temperatures would allow a reduction in radiator size. Terrestrial applications include deep-well drilling instrumentation, high power electronics, and nuclear reactor instrumentation and control. To meet the needs of the applications mentioned previously, the high-temperature electronics (HTE) program at the Lewis Research Center is developing silicon carbide (SiC) as a high-temperature semiconductor material. Research is focused on developing the crystal growth, growth modeling, characterization, and device fabrication technologies necessary to produce a family of SiC devices. Interest in SiC has grown dramatically in recent years due to solid advances in the technology. Much research remains to be performed, but SiC appears ready to emerge as a useful semiconductor material.

  8. Highly compliant transparent electrodes

    NASA Astrophysics Data System (ADS)

    Shian, Samuel; Diebold, Roger M.; McNamara, Alena; Clarke, David R.

    2012-08-01

    Adaptive optical devices based on electric field induced deformation of dielectric elastomers require transparent and highly compliant electrodes to conform to large shape changes. Electrical, optical, and actuation properties of acrylic elastomer electrodes fabricated with single-walled carbon nanotubes (SWCNTs) and silver nanowires (AgNWs) have been evaluated. Based on these properties, a figure of merit is introduced for evaluating the overall performance of deformable transparent electrodes. This clearly indicates that SWCNTs outperform AgNWs. Under optimal conditions, optical transparency as high as 91% at 190% maximum actuation strain is readily achievable using SWCNT electrodes.

  9. High strength alloys

    SciTech Connect

    Maziasz, Phillip James; Shingledecker, John Paul; Santella, Michael Leonard; Schneibel, Joachim Hugo; Sikka, Vinod Kumar; Vinegar, Harold J.; John, Randy Carl; Kim, Dong Sub

    2012-06-05

    High strength metal alloys are described herein. At least one composition of a metal alloy includes chromium, nickel, copper, manganese, silicon, niobium, tungsten and iron. System, methods, and heaters that include the high strength metal alloys are described herein. At least one heater system may include a canister at least partially made from material containing at least one of the metal alloys. At least one system for heating a subterranean formation may include a tublar that is at least partially made from a material containing at least one of the metal alloys.

  10. High strength alloys

    SciTech Connect

    Maziasz, Phillip James; Shingledecker, John Paul; Santella, Michael Leonard; Schneibel, Joachim Hugo; Sikka, Vinod Kumar; Vinegar, Harold J; John, Randy Carl; Kim, Dong Sub

    2010-08-31

    High strength metal alloys are described herein. At least one composition of a metal alloy includes chromium, nickel, copper, manganese, silicon, niobium, tungsten and iron. System, methods, and heaters that include the high strength metal alloys are described herein. At least one heater system may include a canister at least partially made from material containing at least one of the metal alloys. At least one system for heating a subterranean formation may include a tubular that is at least partially made from a material containing at least one of the metal alloys.

  11. Delirium at high altitude.

    PubMed

    Basnyat, Buddha

    2002-01-01

    A 35-year-old man on a trek to the Mount Everest region of Nepal presented with a sudden, acute confusional state at an altitude of about 5000 m. Although described at higher altitudes, delirium presenting alone has not been documented at 5000 m or at lower high altitudes. The differential diagnosis which includes acute mountain sickness and high altitude cerebral edema is discussed. Finally, the importance of travelling with a reliable partner and using proper insurance is emphasized in treks to the Himalayas. PMID:12006167

  12. High intensity hadron accelerators

    SciTech Connect

    Teng, L.C.

    1989-05-01

    This rapporteur report consists mainly of two parts. Part I is an abridged review of the status of all High Intensity Hadron Accelerator projects in the world in semi-tabulated form for quick reference and comparison. Part II is a brief discussion of the salient features of the different technologies involved. The discussion is based mainly on my personal experiences and opinions, tempered, I hope, by the discussions I participated in in the various parallel sessions of the workshop. In addition, appended at the end is my evaluation and expression of the merits of high intensity hadron accelerators as research facilities for nuclear and particle physics.

  13. High Brightness Test Stand

    SciTech Connect

    Birx, D.L.; Caporaso, G.J.; Boyd, J.K.; Hawkins, S.A.; Poor, S.E.; Reginato, L.L.; Rogers, D. Jr.; Smith, M.W.

    1985-08-07

    The High Brightness Test Stand is a 2 MeV, less than or equal to 10 kA electron accelerator module. This accelerator module, designed as an upgrade prototype for the Advanced Test Accelerator (ATA), combines solid state nonlinear magnetic drives with state-of-the-art induction linac technology. The facility serves a dual role, as it not only provides a test bed for this new technology, but is used to develop high brightness electron optics. We will both further describe the accelerator, as well as present some of the preliminary electron optics measurements.

  14. High Energy Astrophysics Program

    NASA Technical Reports Server (NTRS)

    1996-01-01

    This report reviews activities performed by members of the USRA (Universities Space Research Association) contract team during the six months during the reporting period (10/95 - 3/96) and projected activities during the coming six months. Activities take place at the Goddard Space Flight Center, within the Laboratory for High Energy Astrophysics. Developments concern instrumentation, observation, data analysis, and theoretical work in Astrophysics. Missions supported include: Advanced Satellite for Cosmology and Astrophysics (ASCA), X-ray Timing Experiment (XTE), X-ray Spectrometer (XRS), Astro-E, High Energy Astrophysics Science, Archive Research Center (HEASARC), and others.

  15. High Energy Astrophysics Program

    NASA Technical Reports Server (NTRS)

    1996-01-01

    This report reviews activities performed-by members of the USRA contract team during the six months of the reporting period and projected activities during the coming six months. Activities take place at the Goddard Space Flight Center, visiting the Laboratory for High Energy Astrophysics. Developments concern instrumentation, observation, data analysis, and theoretical work in Astrophysics. Missions supported include: Advanced Satellite for Cosmology and Astrophysics (ASCA); X-ray Timing Experiment (XTE); X-ray Spectrometer (XRS); Astro-E; High Energy Astrophysics Science Archive Research Center (HEASARC), and others.

  16. High Performance FORTRAN

    NASA Technical Reports Server (NTRS)

    Mehrotra, Piyush

    1994-01-01

    High performance FORTRAN is a set of extensions for FORTRAN 90 designed to allow specification of data parallel algorithms. The programmer annotates the program with distribution directives to specify the desired layout of data. The underlying programming model provides a global name space and a single thread of control. Explicitly parallel constructs allow the expression of fairly controlled forms of parallelism in particular data parallelism. Thus the code is specified in a high level portable manner with no explicit tasking or communication statements. The goal is to allow architecture specific compilers to generate efficient code for a wide variety of architectures including SIMD, MIMD shared and distributed memory machines.

  17. High Availability Electronics Standards

    SciTech Connect

    Larsen, R.S.; /SLAC

    2006-12-13

    Availability modeling of the proposed International Linear Collider (ILC) predicts unacceptably low uptime with current electronics systems designs. High Availability (HA) analysis is being used as a guideline for all major machine systems including sources, utilities, cryogenics, magnets, power supplies, instrumentation and controls. R&D teams are seeking to achieve total machine high availability with nominal impact on system cost. The focus of this paper is the investigation of commercial standard HA architectures and packaging for Accelerator Controls and Instrumentation. Application of HA design principles to power systems and detector instrumentation are also discussed.

  18. High Median Nerve Injuries.

    PubMed

    Isaacs, Jonathan; Ugwu-Oju, Obinna

    2016-08-01

    The median nerve serves a crucial role in extrinsic and intrinsic motor and sensory function to the radial half of the hand. High median nerve injuries, defined as injuries proximal to the anterior interosseous nerve origin, therefore typically result in significant functional loss prompting aggressive surgical management. Even with appropriate recognition and contemporary nerve reconstruction, however, motor and sensory recovery may be inadequate. With isolated persistent high median nerve palsies, a variety of available tendon transfers can improve key motor functions and salvage acceptable use of the hand. PMID:27387077

  19. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    SciTech Connect

    Del Cul, G.D.

    2002-10-01

    The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the technologies used in

  20. ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

    SciTech Connect

    Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

    2009-02-23

    component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.