Sample records for core burnup calculations

  1. 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-section data for burnup calculations, using some of the main available evaluated nuclear data files (ENDF-B-VI-Rel.8, JEFF-3.0, JENDL-3.3), on an isotope-by-isotope basis as much as possible. The selected experimental burnup benchmarks are reference cases for LWR and HWR reactors, with analysis of isotopic composition as a function of burnup. For LWR (H2O-moderated uranium oxide lattices) four benchmarks are included: ATM-104 NEA Burnup credit criticality benchmark; Yankee-Rowe Core V; H.B.Robinson Unit 2 and Turkey Point Unit 3. For HWR (D2O-moderated uranium oxide cluster lattices), three benchmarks were selected: NPD-19-rod Fuel Clusters; Pickering-28-rod Fuel Clusters; and Bruce-37-rod Fuel Clusters. The isotopes with experimental concentration data included in these benchmarks are: Se-79, Sr90, Tc99, Ru106, Sn126, Sb125,1129, Cs133-137, Nd143, 145, Sm149-150, 152, Eul53-155, U234-235, 238, Np237, Pu238-242, Am241-243, and Cm242-248. Results and analysis of differences between calculated and measured absolute and/or relative concentrations of these isotopes for the seven benchmarks are included in this work.

  2. Power excursion analysis for high burnup cores

    SciTech Connect

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

    1996-02-01

    A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other transients in which there is a power excursion. The last objective was to investigate the sources of uncertainty in the RDA analysis. In a boiling water reactor, the events identified as having significant energy deposition in the fuel were a rod drop accident, a recirculation flow control failure, and the overpressure events; in a pressurized water reactor, they were a rod ejection accident and boron dilution events. The RDA analysis was done with RAMONA-4B, a computer code that models the space- dependent neutron kinetics throughout the core along with the thermal hydraulics in the core, vessel, and steamline. The results showed that the calculated maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important uncertainties in each of these categories are discussed in the report.

  3. TRIGA fuel burn-up calculations and its confirmation

    Microsoft Academic Search

    R. Khan; S. Karimzadeh; H. Böck

    2010-01-01

    The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and

  4. Calculation of burnup of a black neutron absorber

    SciTech Connect

    Yudkevich, M. S., E-mail: umark@adis.vver.kiae.ru [Russian Research Centre Kurchatov Institute (Russian Federation)

    2011-12-15

    The procedure of calculation of burnup of fuel and strong neutron absorber in a nuclear reactor is described. The method proposed here makes it possible to avoid difficulties associated with heterogeneous blocking of the absorption cross section. The effectiveness of the method is demonstrated by an example.

  5. 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.

  6. Design study for an advanced liquid-metal fast breeder reactor core with a high burnup

    Microsoft Academic Search

    T. Inagaki; H. Kuga; M. Suzuki; T. Yokoyama; M. Yamaoka; K. Kaneto; M. Ohashi; K. Kurihara

    1989-01-01

    Design studies are performed for a commercial liquid-metal fast breeder reactor core that can achieve a burnup of 200 GWd\\/t. A plutonium-type asymmetric parfait core with two different plutonium-enriched zones in the axial direction as well as in the radial direction is studied. This core concept solves core design problems related to high burnup, and it is possible to achieve

  7. 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. [Forschungszentrum Karlsruhe, Institut for Reactor Safety, Hermann-von-Helmholtz-Platz-1, 76344 Eggenstein-Leopoldshafen (Germany)

    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)

  8. 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.

  9. Approach to equilibrium fuelling scheme of 500 MWe PFBR based on 3-D core burnup modeling

    Microsoft Academic Search

    K. Devan; A. Riyas; P. Mohanakrishnan

    2011-01-01

    Approach to equilibrium fuelling scheme of 500MWe prototype fast breeder reactor (PFBR) has been predicted using detailed 3-D core burnup modeling. Equilibrium is reached after two cycles of 180 effective full power days (efpd) each. One-third core is refueled every time in a repeatable scatter load scheme after every 3 cycles. Considering the constraints of linear heat rating (LHR) on

  10. Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR

    Microsoft Academic Search

    C. L. Cowan; R. Protsik; J. W. Lewellen

    1984-01-01

    The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the

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

    NASA Astrophysics Data System (ADS)

    Amiri, Benjamin W.; Poston, David I.

    2002-01-01

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

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

    SciTech Connect

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

    2012-01-01

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

  13. LASER; spectrum calculations with burnup cylindrical lattice. [IBM360,370; FORTRAN IV

    Microsoft Academic Search

    2008-01-01

    LASER is based on modified versions of the slowing-down program MUFT and the thermalization transport theory program THERMOS, and performs a calculation of the neutron spectrum in a uniform lattice made up of cylindrical rods, cladding, and surrounding moderator. The thermal cutoff in LASER is 1.855 eV. The program performs a burnup calculation for the lattice. The spatial distribution of

  14. 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. [Tohoku Univ., Graduate School of Engineering, Dept. of Quantum Science and Energy Engineering, Sendai 980-8579 (Japan); Suyama, K. [Japan Atomic Energy Agency, Shirakata-Shirane 2-4, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    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)

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

    SciTech Connect

    Karpushkin, T. Yu., E-mail: timka83@yandex.ru [Russian Research Centre Kurchatov Institute (Russian Federation)

    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.

  16. 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.

  17. Environment-based pin-power reconstruction method for homogeneous core calculations

    SciTech Connect

    Leroyer, H.; Brosselard, C.; Girardi, E. [EDF R and D/SINETICS, 1 av du General de Gaulle, F92141 Claman Cedex (France)

    2012-07-01

    Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOX assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)

  18. IN-CORE FUEL MANAGEMENT: PWR Core Calculations Using MCRAC

    NASA Astrophysics Data System (ADS)

    Petrovi?, B. G.

    1991-01-01

    The following sections are included: * INTRODUCTION * IN-CORE FUEL MANAGEMENT CALCULATIONS * In-Core Fuel Management * Methodological Problems of In-Core Fuel Management * In-Core Fuel Management Analytical Tools * PENN STATE FUEL MANAGEMENT PACKAGE * Penn State Fuel Management Package (PFMP) * Assembly Data Description (ADD) * Linking PSU-LEOPARD and MCRAC: An Example * MULTICYCLE REACTOR ANALYSIS CODE (MCRAC) * Main Features and Options of MCRAC code * Core geometry * Diffusion equations * 1.5-group model * Multicycle neutronic analysis * Multicycle cost analysis * Criticality search * Power-dependent xenon feedback calculations * Control rod and burnable absorber simulation * Search for LP with flat BOC power distribution * Artificial ADD option * Variable dimensioning technique * RBI version of MCRAC code * Programming changes in PC version * Fuel interchange option * MCRAC Input/Output * General input description * Sample input * Sample output * EXPERIENCE WITH MCRAC CODE * CONCLUSIONS * REFERENCES

  19. Fluence-limited burnup as a function of fast reactor core parameters

    E-print Network

    Kersting, Alyssa (Alyssa Rae)

    2011-01-01

    The limiting factor in current designs for fast reactors is not only the reactivity, but also the maximum permissible fast-neutron fluence in the cladding, especially for reduced uranium enrichment cores using high-albedo ...

  20. Improvements in EBR-2 core depletion calculations

    SciTech Connect

    Finck, P.J.; Hill, R.N.; Sakamoto, S.

    1991-01-01

    The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs.

  1. Demonstration of a high burnup heterogeneous core using ferritic/martensitic materials

    SciTech Connect

    Lovell, A.J.; Fox, G.L.; Sutherland, W.H.; Hecht, S.L.

    1986-04-11

    The purpose of the Core Demonstration Experiment (CDE) is to demonstrate the capability of a mixed-oxide fuel system to achieve a three year life in a prototypic LMR heterogeneous reactor environment. The CDE assemblies are fabricated using wire-wrapped, large-diameter, advanced-oxide fuel and blanket pins with tempered martensitic HT9 cladding, wire wrap, and duct. The highest power fuel assembly operates with a Beginning of Life (BOL) peak linear pin power of 445 W/cm and a peak cladding temperature of 615C. The fuel and blanket assembly irradiation will start in FFTF Cycle 9 and continue for about 900 Equivalent Full Power Days (EFPD). The successful utilization of the tempered martensitic HT9 alloy in an FFTF test assembly is fully anticipated. The low swelling, observed at intermediate neutron fluence and projected to higher fluences, together with reasonable creep behavior gives acceptable mechanical performance for fuel pins, blanket pins and ducts. Duct length increase, dilation and bow; plus fuel and blanket pin diameter increases remain within specified tolerances. In addition, stress rupture data from unirradiated HT9 imply cumulative damage fractions for the nominal fuel and blanket pins that are low.

  2. 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.

  3. Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

    SciTech Connect

    Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

    2006-07-01

    Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

  4. Dose Rate Calculations for Rotary Mode Core Sampling Exhauster

    SciTech Connect

    FOUST, D.J.

    2000-10-26

    This document provides the calculated estimated dose rates for three external locations on the Rotary Mode Core Sampling (RMCS) exhauster HEPA filter housing, per the request of Characterization Field Engineering.

  5. 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).

  6. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Butler, C.; Albright, D.

    2007-01-01

    Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

  7. 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).

  8. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    SciTech Connect

    DOE

    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 and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

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

    SciTech Connect

    Ariani, Menik [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Su'ud, Zaki; Waris, Abdul; Asiah, Nur [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Shafii, M. Ali [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Andalas University, Kampus Limau Manis, Padang, Sumatera Barat (Indonesia); 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.

  10. 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.

  11. Bethe-Salpeter equation calculations of core excitation spectra

    NASA Astrophysics Data System (ADS)

    Vinson, J.; Rehr, J. J.; Kas, J. J.; Shirley, E. L.

    2011-03-01

    We present a hybrid approach for Bethe-Salpeter equation (BSE) calculations of core excitation spectra, including x-ray absorption (XAS), electron energy loss spectra (EELS), and nonresonant inelastic x-ray scattering (NRIXS). The method is based on ab initio wave functions from the plane-wave pseudopotential code abinit; atomic core-level states and projector augmented wave (PAW) transition matrix elements; the NIST core-level BSE solver; and a many-pole self-energy model to account for final-state broadening and self-energy shifts. Multiplet effects are also approximately accounted for. The approach is implemented using an interface dubbed OCEAN (Obtaining Core Excitations using abinit and NBSE). To demonstrate the utility of the code we present results for the K edges in LiF as probed by XAS and NRIXS, the K edges of KCl as probed by XAS, the Ti L2,3 edge in SrTiO3 as probed by XAS, and the Mg L2,3 edge in MgO as probed by XAS. These results are compared with experiment and with other theoretical approaches.

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

    SciTech Connect

    Irwanto, Dwi [Department of Nuclear Engineering, Tokyo Institute of Technology (Japan); Kato, Yukikata; Obara, Toru [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan); Yamanaka, Ichiro [Department of Applied Chemistry, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

    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.

  13. High Burnup Effects Program

    SciTech Connect

    Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D.

    1990-04-01

    This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the course of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs.

  14. Recent Developments in No-Core Shell-Model Calculations

    SciTech Connect

    Navratil, P; Quaglioni, S; Stetcu, I; Barrett, B R

    2009-03-20

    We present an overview of recent results and developments of the no-core shell model (NCSM), an ab initio approach to the nuclear many-body problem for light nuclei. In this aproach, we start from realistic two-nucleon or two- plus three-nucleon interactions. Many-body calculations are performed using a finite harmonic-oscillator (HO) basis. To facilitate convergence for realistic inter-nucleon interactions that generate strong short-range correlations, we derive effective interactions by unitary transformations that are tailored to the HO basis truncation. For soft realistic interactions this might not be necessary. If that is the case, the NCSM calculations are variational. In either case, the ab initio NCSM preserves translational invariance of the nuclear many-body problem. In this review, we, in particular, highlight results obtained with the chiral two- plus three-nucleon interactions. We discuss efforts to extend the applicability of the NCSM to heavier nuclei and larger model spaces using importance-truncation schemes and/or use of effective interactions with a core. We outline an extension of the ab initio NCSM to the description of nuclear reactions by the resonating group method technique. A future direction of the approach, the ab initio NCSM with continuum, which will provide a complete description of nuclei as open systems with coupling of bound and continuum states is given in the concluding part of the review.

  15. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

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

    SciTech Connect

    Yasunori Ohoka; Ismile; Hiroshi Sekimoto [Tokyo Institute of Technology, Research Laboratory for Nuclear Reactors, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    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)

  17. Calculation of core loss in a novel transformer design

    Microsoft Academic Search

    W. L. Collett; H. R. Buswell

    2011-01-01

    Recent efforts to develop a new power transformer incorporating a novel wire core configuration have required that core magnetic flux and power loss be estimated during the development process. In this work, an innovative core concept was considered using commercial finite element simulation software, with core loss results shown to be comparable to measurements on a standard 10 kVA design.

  18. Transition-phase calculation of a large, heterogeneous-core LMFBR. [SIMMER-II calculations

    SciTech Connect

    Luck, L.B.; Bell, C.R.; Asprey, M.W.; DeVault, G.P.

    1981-01-01

    A mechanistic calculation of a complete transition-phase sequence for a large heterogeneous core LMFBR has been performed using SIMMER-II. Recriticalities occurred as the disruption progressed through a series of different subphases. The number and severity of recriticalities was directly related to the timing and scale of fuel removal and coherence of material motion. The energetics associated with transition-phase are not yet resolved but the understanding of the characteristics of disruption and the effects of uncertainties has been extended significantly.

  19. 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 measured and calculated values for most of the analysed isotopes, similar to those reported previously for lower burnup ranges. Thus, it could be concluded, that SAS2H results for high burnup samples are not subject to higher uncertainty and/or different biases than for lower burnup samples, and that the different isotopic experimental measurement methods provide accurate results with acceptable precision.

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

    SciTech Connect

    Qvist, S.; Reenspan, E. [Dept. of Nuclear Engineering, Univ. of California, Berkeley, CA 94720-1730 (United States)

    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)

  1. Advances in core loss calculations for magnetic materials

    NASA Technical Reports Server (NTRS)

    Triner, J. E.

    1982-01-01

    A new analytical technique which predicts the basic magnetic properties under various operating conditions encountered in state-of-the-art dc-ac/dc converters is discussed. Using a new flux-controlled core excitation circuit, magnetic core characteristics were developed for constant values of ramp flux (square wave voltage excitation) and frequency. From this empirical data, a mathematical loss characteristics equation is developed to analytically predict the specific core loss of several magnetic materials under various waveform excitation conditions. In addition, these characteristics show the circuit designer for the first time the direct functional relatonships between induction level and specific core loss as a function of the two key dc-dc converter operating parameters of input voltage and duty cycle.

  2. Calculation methods for core distortions and mechanical behavior

    SciTech Connect

    Sutherland, W.H.

    1984-09-01

    This paper describes ABADAN, a general purpose, nonlinear, multi-dimensional finite element structural analyses computer code developed for the express purpose of solving large nonlinear problems as typified by the Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System design problem. All of the structural modeling features inherent in a general purpose finite element code and required to adequately model an LMFBR core restraint system are demonstrated. Typical results for a radial row and a sixty degree sector model of FFTF are presented. The sixty degree sector results are interpreted in terms of the design criteria that the core restraint system must satisfy. Extensions and adaptations of these modeling techniques to different core restraint design concepts can be readily achieved. 27 figures.

  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 [Nuclear Research Center National University of Mongolia Ulaanbaatar (Mongolia)

    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. 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.

  6. A PWR Thorium Pin Cell Burnup Benchmark

    SciTech Connect

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

    2000-05-01

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

  7. Coke burnup in a dry coke quenching device and methods of decreasing it

    SciTech Connect

    Filonenko, Yu.Ya.; Naumkin, V.A.; Eremenko, G.N.; Sobolev, S.Ya.; Rusakov, Yu.V.

    1984-01-01

    Decreasing coke burnup in a coke dry quenching device is one method of improving the technicoeconomic indices of its functioning. Unfortunately, at present there is no standard method of calculating coke burnup. This prevents a thorough analysis to be made of the efficiency of the functioning of either individual dry coke quenching device (DCQD) chambers or the devices in general. Coke burnup in a DCQD can be calculated from the equation for the thermal balance of the system. The calculation method is described.

  8. Isotopic Bias and Uncertainty for Burnup Credit Applications

    SciTech Connect

    J.M. Scaglione

    2002-08-19

    The application of burnup credit requires calculating the isotopic inventory of the irradiated fuel. The depletion calculation simulates the burnup of the fuel under reactor operating conditions. The result of the depletion analysis is the predicted isotopic composition, which is ultimately input to a criticality analysis to determine the system multiplication factor (k{sub eff}). This paper demonstrates an approach for calculating the isotopic bias and uncertainty in k{sub eff} for commercial spent nuclear fuel burnup credit. This paper covers 74 different radiochemical assayed spent fuel samples from 22 different fuel assemblies that were irradiated in eight different pressurized water reactors (PWRs). The samples evaluated span an enrichment range of 2.556 wt% U-235 through 4.67 wt% U-235, and burnups from 6.92 GWd/MTU through 55.7 GWd/MTU.

  9. RMC - A Monte Carlo Code for Reactor Core Analysis

    NASA Astrophysics Data System (ADS)

    Wang, Kan; Li, Zeguang; She, Ding; Liang, Jin'gang; Xu, Qi; Qiu, Yishu; Yu, Jiankai; Sun, Jialong; Fan, Xiao; Yu, Ganglin

    2014-06-01

    A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor core analysis on high-performance computing platforms. To meet the requirements of reactor analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the effciency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor analysis problems with good performances.

  10. A Parallel Full Core Transport Calculation Based On Domain Decomposition Method

    NASA Astrophysics Data System (ADS)

    Lenain, Roland; Masiello, Emiliano; Sanchez, Richard; Damian, Frederic

    2014-06-01

    A new interactive homogenization procedure for reactor core or colorset calculations is proposed that requires iterative transport assembly and diffusion core calculations. At each iteration the transport solution of every assembly is used to produce homogenized cross sections for the core calculation. The converged solution gives assembly fine multigroup transport fluxes that preserve macrogroup assembly exchanges in the core. This homogenization avoids the periodic lattice - leakage model approximation and gives detailed assembly transport fluxes without need for an approximated flux reconstruction. In this paper we combined the benefit of a Domain Decomposition Method, that split the original transport problem in several multigroup fixed-source problems, with the effective solution of the Coarse-Mesh Finite-Differences operator that provides the whole-core eigenvalue and the neutron exchange between assemblies.

  11. /sup 18/O as a core plus two valence neutrons: A three-body Faddeev calculation

    SciTech Connect

    Ueta, K.; Miyake, H.; Mizukami, A.

    1983-01-01

    The nucleus /sup 18/O is studied assuming a three-body model: two neutrons outside an inert core of /sup 16/O: and solving the Faddeev equations. The calculated spectrum is in good agreement with experiment.

  12. New Soft-Core Potential Function for Molecular Dynamics Based Alchemical Free Energy Calculations

    E-print Network

    de Groot, Bert

    New Soft-Core Potential Function for Molecular Dynamics Based Alchemical Free Energy Calculations require the modification of the classical nonbonded potential energy terms by applying soft-core potential functions to avoid singularity points. In this work, we propose a novel formulation for a soft

  13. High-burnup oxide fuel in European fast reactors

    Microsoft Academic Search

    K. M. Swanson; A. Languille; G. Muhling

    1989-01-01

    The European Collaboration on Fast Reactors is working on the design of a common demonstrate fast reactor, the European Fast Reactor (EFR) designed to be licensable in all the countries of the collaboration. The first consistent design of EFR calls for uranium-plutonium-oxide fuel assemblies. The first core target is a peak burnup of 15 at.% at a neutron displacement dose

  14. Designing Critical Experiments in Support of Full Burnup Credit

    SciTech Connect

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

    2008-01-01

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

  15. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    SciTech Connect

    Hartini, Entin, E-mail: entin@batan.go.id; Andiwijayakusuma, Dinan, E-mail: entin@batan.go.id [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)

    2014-09-30

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  16. 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.

  17. An improved energy-collapsing method for core-reflector modelization in SFR core calculations using the PARIS platform

    SciTech Connect

    Vidal, J. F.; Archier, P.; Calloo, A.; Jacquet, P.; Tommasi, J. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-lez-Durance (France); Le Tellier, R. [CEA, DEN, DTN, Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2012-07-01

    In the framework of the ASTRID project, sodium cooled fast reactor studies are conducted at CEA in compliance with GEN IV reactors criteria, particularly for safety requirements. An improved safety requires better calculation tools to obtain accurate reactivity effects (especially sodium void effect) and power map distributions. The current calculation route lies on the JEFF3.1.1 library and the classical two-step approach performed with the ECCO module of the ERANOS code system at the assembly level and the Sn SNATCH solver - implemented within the PARIS platform - at the core level. 33-group cross sections used by SNATCH are collapsed from 1968-group self-shielded cross-section with a specific flux-current weighting. Recent studies have shown that this collapsing is non-conservative when dealing with core-reflector interface and can lead to reactivity discrepancies larger than 500 pcm in the case of a steel reflector. Such a discrepancy is due to the flux anisotropy at the interface, which is not taken into account when cross sections are obtained from separate fuel and reflector assembly calculations. A new approach is proposed in this paper. It consists in separating the self-shielding and the flux calculations. The first one is still performed with ECCO on separate patterns. The second one is done with SNATCH on a 1D traverse, representative of the core-reflector interface. An improved collapsing method using angular flux moments is then carried out to collapse the cross sections onto the 33-group structure. In the case of a simplified ZONA2B 2D homogeneous benchmark, results in terms of k{sub eff} and power map are strongly improved for a small increase of the computing time. (authors)

  18. Effect of burnup on ACR-700 3-D reactivity devices cross sections

    SciTech Connect

    Dahmani, M.; Marleau, G.; Varin, E. [Institut de Genie Nucleaire, Ecole Polytechnique de Montreal, 2900 Boulevard Edouard-Montpetit, Montreal, Que. H3T 1J4 (Canada)

    2006-07-01

    Full core analysis of typical power reactors being generally performed using few groups diffusion theory, it is necessary to generate beforehand, using a lattice code, the required few group cross sections and diffusion coefficients associated with each region in the core. For CANDU-type reactors including the Advanced CANDU Reactor (ACR), the problem is more complex because these reactors contain vertical reactivity devices that are located between two horizontal fuel bundles. The usual calculation scheme relies in this case on a 2-D fuel cell calculation to generate the few group fuel properties and on a 3-D supercell calculation for the analysis of the reactivity devices present in the core. Because of its complexity, the supercell calculations are generally performed using simplified fuel geometries. In this paper, the different stages involved in the reactor physics simulations for ACR will be explained focusing particularly on a study of the burnup dependence of the incremental cross section associated with zone control units (ZCU). The use of these incremental cross sections for finite core calculations will also be presented. (authors)

  19. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

    SciTech Connect

    Ellis, RJ

    2001-06-01

    The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

  20. Validation of finite difference core diffusion calculation methods with FEM and NEM for VVER-1000 MWe reactor

    SciTech Connect

    Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); RPDD, Central Complex, BARC, Mumbai - 400085 (India); Singh, T. [Reactor Physics and Nuclear Engineering Section, Reactor Group, BARC, Mumbai (India); Pal, U.; Karthikeyan, R. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); Sundaram, G. [Nuclear Safety Group, KK-NPC, Mumbai (India)

    2006-07-01

    India is developing several in-house fuel management codes for the design evaluation of WER-1000 M We reactors, being built at Kudankulam, Tamil Nadu in collaboration with Russian Federation. A lattice burnup code EXCEL provides the few group lattice parameters of various fuel assembly types constituting the core. The core diffusion analyses have been performed by two methods. In the first method the entire fuel assembly is treated as a single homogenized cell. Each fuel assembly cell is divided into 6n{sup 2} triangles, where 'n' is the number of uniform divisions on a side of the hexagon. Regular triangular meshes are used in the active core as well as in surrounding reflector regions. This method is incorporated in the code TRIHEXFA. In the second method a pin by pin description of the core is accomplished by considering the few group lattice parameters generated by EXCEL code for various fuel and non-fuel cells in each fuel assembly. Regular hexagonal cells of one pin pitch are considered in the core and reflector regions. This method is incorporated in HEXPIN code. Both these codes use centre mesh finite difference method (FDM) for regular triangular or hexagonal meshes. It is well known that the large size of the WER fuel assembly, the zigzag structure of the core-baffle zone, the distribution of water tubes of different diameter in this baffle zone and the surrounding steel and water layers of different thickness, all lead to a very complex description of the core-reflector interface. We are analyzing the WER core in fresh state by two other approaches to obtain independent benchmark reference solutions. They are finite element method (FEM) and nodal expansion method (NEM). The few group cross sections of EXCEL are used in the FEM and NEM analyses. The paper would present the comparison of the results of core followup simulations of FD codes with those of FEM and NEM analyses. (authors)

  1. Importance of nonlinear core corrections for density-functional based pseudopotential calculations

    E-print Network

    Liu, Amy Y.

    -functional based pseudopotential PSP calculations. The quality of the PSP approach is assessed by comparing, one can construct pseudopotentials PSP's which effectively project out the core states from of work aimed at mesh-based approaches.23­28 Nevertheless, many PSP applications use plane waves. The size

  2. An improved resonance self-shielding method for heterogeneous fast reactor assembly and core calculations

    SciTech Connect

    Lee, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4842 (United States); Yang, W. S. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907-2017 (United States)

    2013-07-01

    An improved resonance self-shielding method has been developed to accurately estimate the effective multigroup cross sections for heterogeneous fast reactor assembly and core calculations. In the method, the heterogeneity effect is considered by the use of isotopic escape cross sections while the resonance interference effect is accounted for through the narrow resonance approximation or slowing-down calculations for specific compositions. The isotopic escape cross sections are calculated by solving fixed-source transport equations with the method of characteristics for the whole problem domain. This method requires no pre-calculated resonance integral tables or parameters that are typically necessary in the subgroup method. Preliminary results for multi pin-cell fast reactor problems show that the escape cross sections estimated from the explicit-geometry fixed source calculations produce more accurate eigenvalue and self-shielded effective cross sections than those from conventional one-dimensional geometry models. (authors)

  3. Full Core 3-D Simulation of a Partial MOX LWR Core

    SciTech Connect

    S. Bays; W. Skerjanc; M. Pope

    2009-05-01

    A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch average discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.

  4. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

    PubMed

    Kurosawa, Masahiko

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. PMID:16604689

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

    SciTech Connect

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this case, a self-calibration method was developed to obtain the spectrometer's relative efficiency curve based upon gamma lines emitted from {sup 140}La. It was found that the ratio of {sup 239}Np/{sup 132}I can be used in burnup measurement with an uncertainty of {approx} {+-}3% throughout the pebble's lifetime. In addition, by doping the fuel with {sup 60}Co, the use of the {sup 60}Co/{sup 134}Cs and {sup 239}Np/{sup 132}I ratios can simultaneously yield the enrichment and burnup of each pebble. A functional gamma-ray spectrometry measurement system was constructed and tested with light water reactor fuels. Experimental results were observed to be consistent with the predictions. On using the passive neutron counting method for the on-line burnup measurement, it was found that neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged cross sections used in the depletion calculations; thus a large uncertainty exists in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and neutron emission rate is too low so that it is impossible to determine a pebble's burnup by on-line neutron counting. At high burnup levels, due to the decreasing of the uncertainty in neutron emission rate and the super-linear feature of the correlation, the uncertainty in burnup determination was found to be {approx}7% at the discharge burnup, which is acceptable for determining whether a pebble should be discharged or not. In terms of neutron detection, because an irradiated pebble is a weak neutron source and a much stronger gamma source, neutron detector system should have high neutron detection efficiency and strong gamma discrimination capability. Of all the commonly used neutron detectors, the He-3 and BF3 detector systems were found to be able to satisfy the requirement on detection efficiency; but their gamma discrimination capability is only marginal for this on-line application. Even with thick gamma shielding, these two types of detectors sha

  6. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

    SciTech Connect

    Ellis, RJ

    2001-02-02

    The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

  7. Ab initio calculation of core-valence-valence Auger spectra in closed shell systems

    Microsoft Academic Search

    Gian Paolo Brivio; Guido Fratesi; Mario Italo Trioni; Simona Ugenti; Enrico Perfetto; Michele Cini

    2009-01-01

    We propose an ab initio method to evaluate the core-valence-valence Auger spectrum of systems with filled valence bands. The method is based on the Cini-Sawatzky theory, and aims at estimating the parameters by first-principles calculations in the framework of DFT. Photoemission energies and the interaction energy for the two holes in the final state are evaluated by performing DFT simulations

  8. Ab initio calculation of core-valence-valence Auger spectra in closed shell systems

    Microsoft Academic Search

    G. Fratesi; M. I. Trioni; G. P. Brivio; S. Ugenti; E. Perfetto; M. Cini

    2008-01-01

    We propose an ab initio method to evaluate the core-valence-valence Auger spectrum of systems with filled valence bands. The method is based on the Cini-Sawatzky theory and aims at estimating the parameters by first-principles calculations in the framework of density-functional theory (DFT). Photoemission energies and the interaction energy for the two holes in the final state are evaluated by performing

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

    SciTech Connect

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik [Nuclear Physics and Biophysics Research Division, Physics Department, Institut Teknologi Bandung (Indonesia); Suzuki, Mitsutoshi [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA) (Japan)

    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.

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

    Microsoft Academic Search

    Exarhos

    1985-01-01

    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

  11. Calculation of scattering characteristic of complex target on multi-core platform

    NASA Astrophysics Data System (ADS)

    Guo, Xing; Wu, Zhensen; Linghu, Longxiang

    2013-09-01

    The scattering characteristic of complex target from terrestrial and celestial background radiation has been widely used in such engineering fields as remote sensing, feature extraction, tracking and recognition of target thus having been an attractive field for many scientists for decades. In our method, the model of target is constructed using 3DMAX and the surface is divided into triangle facets firstly. Bidirectional Reflectance Distribution Function (BRDF) is introduced and MODTRAN is applied to calculate background radiation for a given time at a given place. Finally the scattering of each facet is added up to get the scattering of the target. As the background radiance comes in all directions and in a wide spectrum and the complex target always consists of thousands of facets, in general it takes hours to complete the calculation. Consequently this limits its use in the real time applications. Recent years have seen the continual development of multi-core CPU. As a result parallel programming on multi-cores has been more and more popular. In this paper, the openMP, Intel CILK ++, Intel Threading Building Blocks (TBB) are used separately to leverage the processing power of multi-cores processors. Our experiments are conducted on a DELL desktop based on an Intel I7- 2600K CPU running at 3.40 GHz with 8 cores and 16.0 GB RAM. The Intel Composer 2013 is employed to build the program. Also in OpenMP implementation, gcc is used. The results demonstrate that highest speedups for three parallel models are 5.06X, 5.02X, 5.15X respectively.

  12. Development of a Fully-Automated Monte Carlo Burnup Code Monteburns

    SciTech Connect

    Poston, D.I.; Trellue, H.R.

    1999-01-01

    Several computer codes have been developed to perform nuclear burnup calculations over the past few decades. In addition, because of advances in computer technology, it recently has become more desirable to use Monte Carlo techniques for such problems. Monte Carlo techniques generally offer two distinct advantages over discrete ordinate methods: (1) the use of continuous energy cross sections and (2) the ability to model detailed, complex, three-dimensional (3-D) geometries. These advantages allow more accurate burnup results to be obtained, provided that the user possesses the required computing power (which is required for discrete ordinate methods as well). Several linkage codes have been written that combine a Monte Carlo N-particle transport code (such as MCNPTM) with a radioactive decay and burnup code. This paper describes one such code that was written at Los Alamos National Laboratory: monteburns. Monteburns links MCNP with the isotope generation and depletion code ORIGEN2. The basis for the development of monteburns was the need for a fully automated code that could perform accurate burnup (and other) calculations for any 3-D system (accelerator-driven or a full reactor core). Before the initial development of monteburns, a list of desired attributes was made and is given below. o The code should be fully automated (that is, after the input is set up, no further user interaction is required). . The code should allow for the irradiation of several materials concurrently (each material is evaluated collectively in MCNP and burned separately in 0RIGEN2). o The code should allow the transfer of materials (shuffling) between regions in MCNP. . The code should allow any materials to be added or removed before, during, or after each step in an automated fashion. . The code should not require the user to provide input for 0RIGEN2 and should have minimal MCNP input file requirements (other than a working MCNP deck). . The code should be relatively easy to use and not require several complicated input files. All of these features have been developed fully or partially in monteburns, although several improvements have yet to be implemented.

  13. Burnup study for Pakistan Research Reactor1 utilizing high density low enriched uranium fuel

    Microsoft Academic Search

    Rizwan Ahmed; Aslam; Nasir Ahmad

    2005-01-01

    Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel

  14. Emergence of rotational bands in ab initio no-core configuration interaction calculations

    E-print Network

    Caprio, M A; Vary, J P; Smith, R

    2015-01-01

    Rotational bands have been observed to emerge in ab initio no-core configuration interaction (NCCI) calculations for p-shell nuclei, as evidenced by rotational patterns for excitation energies, electromagnetic moments, and electromagnetic transitions. We investigate the ab initio emergence of nuclear rotation in the Be isotopes, focusing on 9Be for illustration, and make use of basis extrapolation methods to obtain ab initio predictions of rotational band parameters for comparison with experiment. We find robust signatures for rotational motion, which reproduce both qualitative and quantitative features of the experimentally observed bands.

  15. Whole-core neutron transport calculations without fuel-coolant homogenization

    SciTech Connect

    Smith, M. A.; Tsoulfanidis, N.; Lewis, E. E.; Palmiotti, G.; Taiwo, T. A.

    2000-02-10

    The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the fill spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.

  16. Calculation of evanescent field interaction with metallic nanoparticles immobilized on the air holes of solid-core photonic crystal fiber

    Microsoft Academic Search

    Alexander Raspopin; Hong-Liang Cui; Henry Du

    2005-01-01

    We report the calculation of the attenuation coefficient of a probing optical mode due to interaction with metallic nanoparticles randomly distributed in the air holes of a solid core photonic crystal fiber (PCF) for SERS-based sensing and detection. The approach employed is an approximation of the solid core PCF with conventional curricular fiber considering a similar total internal reflection mechanism

  17. Hybrid parallel code acceleration methods in full-core reactor physics calculations

    SciTech Connect

    Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)

    2012-07-01

    When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

  18. Calculated coupling efficiency between an elliptical-core optical fiber and an optical waveguide over temperature

    NASA Technical Reports Server (NTRS)

    Tuma, Margaret L.; Weisshaar, Andreas; Li, Jian; Beheim, Glenn

    1995-01-01

    To determine the feasibility of coupling the output of a single-mode optical fiber into a single-mode rib waveguide in a temperature varying environment, a theoretical calculation of the coupling efficiency between the two was investigated. Due to the complex geometry of the rib guide, there is no analytical solution to the wave equation for the guided modes, thus, approximation and/or numerical techniques must be utilized to determine the field patterns of the guide. In this study, three solution methods were used for both the fiber and guide fields; the effective-index method (EIM), Marcatili's approximation, and a Fourier method. These methods were utilized independently to calculate the electric field profile of each component at two temperatures, 20 C and 300 C, representing a nominal and high temperature. Using the electric field profile calculated from each method, the theoretical coupling efficiency between an elliptical-core optical fiber and a rib waveguide was calculated using the overlap integral and the results were compared. It was determined that a high coupling efficiency can be achieved when the two components are aligned. The coupling efficiency was more sensitive to alignment offsets in the y direction than the x, due to the elliptical modal field profile of both components. Changes in the coupling efficiency over temperature were found to be minimal.

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

    SciTech Connect

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

    2009-01-01

    The purpose of this study is to provide insights into the neutronic similarities that may exist between a generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the type of CRC state-points that may be applicable for validation of burnup credit criticality safety calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate system similarity on an integral and nuclide-reaction specific level. The results indicate that, except for the fresh fuel core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to a generic cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. Based on the integral system parameter, C{sub k}, approximately 30 of the 40 CRC state-points are applicable to validation of burnup credit in the generic cask containing typical spent fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. The state-points providing the highest similarity (C{sub k} > 0.95) were attained at or near the end of a reactor cycle. The C{sub k} values are dominated by neutron reactions with major actinides and hydrogen, as the sensitivities of these reactions are much higher than those of the minor actinides and fission products. On a nuclide-reaction specific level, the CRC state-points provide significant similarity for most of the actinides and fission products relevant to burnup credit. A comparison of energy-dependent sensitivity profiles shows a slight shift of the CRC K{sub eff} sensitivity profiles toward higher energies in the thermal region as compared to the K{sub eff} sensitivity profile of the generic cask. Parameters representing coverage of the application by the CRCs on an energy-dependent, nuclide-reaction specific level (i.e., effectiveness of the CRCs for validating the cross sections as used in the application) were also examined. Based on the CRCs with C{sub k} > 0.8 and an assumed relative standard deviation for uncovered covariance data of 25%, the relative standard deviation of K{sub eff} due to uncovered sensitivity data varies from 0.79% to 0.95% for cask burnups ranging from 10 to 60 GWd/MTU. As expected, this uncertainty in K{sub eff} is largely dominated by noncoverage of sensitivities from major actinides and hydrogen. The contributions from fission products and minor actinides are very small and comparable to statistical uncertainties in K{sub eff} results. These results (again, assuming a 25% uncertainty for uncovered covariance data) indicate that there could be approximately 1% uncertainty in the calculated application K{sub eff} due to incomplete neutronic testing (validation) of the software by the CRCs. However, this conclusion also assumes all other uncertainties in the complex CRC configurations (e.g., isotopic compositions of burned fuel, operation history, data) are well known. Thus, an evaluation of the uncertainties in the CRC configurations is needed prior to the use of CRCs for code validation (i.e., quantifying code bias and bias uncertainty).

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

    Microsoft Academic Search

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

    2010-01-01

    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

  1. Testing the improved method for calculating the radiation heat generation at the periphery of the BOR-60 reactor core

    SciTech Connect

    Varivtsev, A. V., E-mail: vav3@niiar.ru; Zhemkov, I. Yu. [JSC “SSC RIAR,” Dimitrovgrad-10 (Russian Federation)

    2014-12-15

    The application of the improved method for calculating the radiation heat generation in the elements of an experimental device located at the periphery of the BOR-60 reactor core results in a significant reduction in the discrepancies between the calculated and the experimental data. This allows us to conclude that the improved method has an advantage over the one used earlier.

  2. Ab initio calculations of the elasticity of hcp-Fe as a function of temperature at inner-core pressure

    E-print Network

    Vocadlo, Lidunka

    initio finite temperature molecular dynamics simulations have been used to calculate the elastic). Knowledge of the elastic constants of iron at core conditions allows determination of a number of propertiesAb initio calculations of the elasticity of hcp-Fe as a function of temperature at inner

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

    SciTech Connect

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

    2011-01-01

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

  4. Properties of metastable alkaline-earth-metal atoms calculated using an accurate effective core potential

    SciTech Connect

    Santra, Robin; Christ, Kevin V.; Greene, Chris H. [Department of Physics and JILA, University of Colorado, Boulder, Colorado 80309-0440 (United States)

    2004-04-01

    The first three electronically excited states in the alkaline-earth-metal atoms magnesium, calcium, and strontium comprise the (nsnp){sup 3}P{sub J}{sup o}(J=0,1,2) fine-structure manifold. All three states are metastable and are of interest for optical atomic clocks as well as for cold-collision physics. An efficient technique--based on a physically motivated potential that models the presence of the ionic core--is employed to solve the Schroedinger equation for the two-electron valence shell. In this way, radiative lifetimes, laser-induced clock shifts, and long-range interaction parameters are calculated for metastable Mg, Ca, and Sr.

  5. Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

    SciTech Connect

    Jung, Y. S.; Joo, H. G. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Yoon, J. I. [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)

    2013-07-01

    The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

  6. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    NASA Astrophysics Data System (ADS)

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

    2014-06-01

    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard deviations and computing times.

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

    Microsoft Academic Search

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

    The TRANSURANUS burn-up model (TUBRNP) calculates the local concentration of the actinides, the main fission products, and 4He as a function of the radial position across a fuel rod. In this paper, the improvements in the helium production model as well as the extensions in the simulation of 238–242Pu, 241Am, 243Am and 242–245Cm isotopes are described. Experimental data used for

  8. Calculs d'assemblages de REP en environnement

    NASA Astrophysics Data System (ADS)

    Leroyer, Hadrien

    Pressurized Water Reactors (PWR) are the most common nuclear reactor used today. The core of a PWR is composed of approximately 200 assemblies immersed in pressurized light water, which can be Uranium Oxyde assemblies (UOX) or Mixed Oxyde assemblies (MOX) coming from the reprocessings of used UOX. Electro-nuclear industries want to calculate the neutron flux inside these reactors, by solving the neutron transport equation, because it controls the dynamic of the core. Actually, the computers' power available today does not allow for a solution to the transport equation over the whole core, in three dimensions, with burnup. This is why reactor physicists use several approximations in order to obtain a solution for the neutron flux. This implies defining a pertinent calculation scheme. Generally, the calculation scheme requires homogenized macroscopic cross sections libraries generated using infinite lattice calculations on assemblies. Several parameters are used for the tabulation of these libraries including the burnup, the temperature and density of the coolant and the fuel, the concentration of boron and of xenon-135. Then the code evaluates the flux distribution in the finite reactor with the diffusion equation using cross sections interpolated from these libraries. However, the infinite lattice hypothesis may not be valid for highly heterogeneous cores, for example a core with burnup gradients or MOX / UOX interfaces. The purpose of this study is to evaluate the physical impact of heterogeneous environment on PWR assemblies. We first define a reference calculation scheme for a 3 x 3 assembly cluster, taking all heterogeneous environment effect into account, with the lattice cell code DRAGON. We later compare this reference with infinite lattice calculations, or with other calculation schemes closer to a full reactor calculation code. Those comparisons will allow us to explain physically the effects of the heterogeneous environment, and also to evaluate the errors in the reactor code committed when this effect is not taken into account. Finally, we will propose solutions to this issue.

  9. Calculated Actinide and Fission Product Concentration Ratios for Gaseous Effluent Monitoring Using Monteburns 3.01

    SciTech Connect

    Charlton, William S. [Los Alamos National Laboratory (United States); Perry, Robert T. [Los Alamos National Laboratory (United States); Fearey, Bryan L. [Los Alamos National Laboratory (United States); Parish, Theodore A. [Texas A and M University (United States)

    2000-08-15

    Techniques have been developed at Los Alamos National Laboratory for accurately calculating certain spent-fuel isotope concentration ratios for pressurized water reactor assemblies using a linked MCNP/ORIGEN2 code named Monteburns 3.01, without resorting to an assembly or full-core calculation. The effects of various fuel parameters such as the number of radial fuel regions per pin, burnup step size, reactor power, reactivity control mechanisms, and axial profiles have been studied. The significance of each factor was determined. A method was also proposed for calculating spent-fuel inventories as a function of burnup for a wide range of reactors and fuel types. It was determined that accurate calculations can be obtained using a three-dimensional, modified pin cell with seven radial fuel regions and two (flat-flux) axial fuel regions calculated with 2000 MWd/tonne U burnup steps for burnups ranging from 0 to 50 000 MWd/tonne U. The calculational technique was benchmarked to measured values from the Calvert Cliffs Unit 1 reactor, and good agreement from the point of view of calibrating a monitoring instrument was found for most cases.

  10. No-core Monte Carlo shell model calculations with unitary correlation operator method and similarity renormalization group

    NASA Astrophysics Data System (ADS)

    Liu, Lang

    2015-05-01

    The unitary correlation operator method (UCOM) and the similarity renormalization group theory (SRG) are compared and discussed in the framework of the no-core Monte Carlo shell model (MCSM) calculations for 3H and 4He. The treatment of spurious center-of-mass motion by Lawson's prescription is performed in the MCSM calculations. These results with both transformed interactions show good suppression of spurious center-of-mass motion with proper Lawson's prescription parameter ?c.m. values. The UCOM potentials obtain faster convergence of total energy for the ground state than that of SRG potentials in the MCSM calculations, which differs from the cases in the no-core shell model calculations (NCSM). These differences are discussed and analyzed in terms of the truncation scheme in the MCSM and NCSM, as well as the properties of the potentials of SRG and UCOM. Supported by Fundamental Research Funds for the Central Universities (JUSRP1035), National Natural Science Foundation of China (11305077)

  11. Edge dislocation core structures in FCC metals determined from ab initio calculations combined with the improved Peierls-Nabarro equation

    NASA Astrophysics Data System (ADS)

    Wang, Rui; Wang, Shaofeng; Wu, Xiaozhi

    2011-04-01

    We have employed the improved Peierls-Nabarro (P-N) equation to study the properties of 1/2lang110rang edge dislocation in the {111} plane in face-centered cubic (FCC) metals Al, Cu, Ir, Pd and Pt. The generalized-stacking-fault energy surface entering the equation is calculated by using first-principles density functional theory (DFT). The accuracy of the method has been tested by calculating the values for various stacking fault energies that favorably compare with previous theoretical and experimental results. The core structures, including the core widths of the edge and screw components, and dissociation behavior have been investigated. The dissociated distance between two partials for Al in our calculation agrees well with the values obtained from numerical simulation with DFT and molecular dynamics simulation, as well as experiment. Our calculations show that it is preferred to create partial dislocations in Cu, and easily observed full dislocations in Al, Ir, Pd and especially Pt.

  12. Ab-initio calculations of core-shell CdSe/ZnS nanowires marcel@physik.tu-berlin.de

    E-print Network

    Nabben, Reinhard

    Ab-initio calculations of core-shell CdSe/ZnS nanowires marcel@physik.tu-berlin.de Calculational) September 2008 M. Mohr, H. Lange and C. Thomsen Structural properties Cross sections of CdSe and CdSe/ZnS 3 Energy[eV] CdSe ZnS Electronic properties VBM CBM 0 2 4 6 nanorod diameter (nm) 0 20 40 60 80 100

  13. All-electron Bethe-Salpeter calculations for shallow-core x-ray absorption near-edge structures

    NASA Astrophysics Data System (ADS)

    Olovsson, W.; Tanaka, I.; Mizoguchi, T.; Puschnig, P.; Ambrosch-Draxl, C.

    2009-01-01

    X-ray absorption near-edge structure spectra are calculated by fully solving the electron/core-hole Bethe-Salpeter equation (BSE) in an all-electron framework. We study transitions from shallow core states, including the Mg L2,3 edge in MgO, the Li K edge in the Li halides LiF, LiCl, LiBr, and LiI, as well as Li2O . We illustrate the advantage of the many-body approach over a core-hole supercell calculation. Both schemes lead to strongly bound excitons, but the nonlocal treatment of the electron-hole interaction in the BSE turns out to be crucial for an agreement with experiment.

  14. 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 predictions. After confirming that the new fuel thermal conductivity model in CTF worked and provided consistent results with FRAPTRAN predictions for a single fuel rod configuration, the same type of analysis was carried out for a bigger system which is the 4x4 PWR bundle consisting of 15 fuel pins and one control guide tube. Steady- state calculations at Hot Full Power (HFP) conditions for control guide tube out (unrodded) were performed using the 4x4 PWR array with CTF/TORT-TD coupled code system. Fuel centerline, surface and average temperatures predicted by CTF/TORT-TD with and without the new fuel thermal conductivity model were compared against CTF/TORT-TD/FRAPTRAN predictions to demonstrate the improvement in fuel centerline predictions when new model was used. In addition to that constant and CTF dynamic gap conductance model were used with the new thermal conductivity model to show the performance of the CTF dynamic gap conductance model and its impact on fuel centerline and surface temperatures. Finally, a Rod Ejection Accident (REA) scenario using the same 4x4 PWR array was run both at Hot Zero Power (HZP) and Hot Full Power (HFP) condition, starting at a position where half of the control rod is inserted. This scenario was run using CTF/TORT-TD coupled code system with and without the new fuel thermal conductivity model. The purpose of this transient analysis was to show the impact of thermal conductivity degradation (TCD) on feedback effects, specifically Doppler Reactivity Coefficient (DRC) and, eventually, total core reactivity.

  15. Electrical and Thermal Conductivity of Liquid Iron at Core Pressures and Temperatures: First-Principles Calculations

    Microsoft Academic Search

    N. de Koker; G. Steinle-Neumann; V. Vlcek

    2010-01-01

    The ability of liquid iron to transport heat and electric charge by conduction at extreme pressure and temperature is of paramount importance to the thermal history of the core. Thermal conductivity determines the amount of heat conducted along the core adiabat, i.e. heat not available for generation of the magnetic field, and also strongly controls the time required for the

  16. 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.

  17. Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

    SciTech Connect

    Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T. [Osaka Univ., 2-1, Yamadaoka, Suita-shi, Osaka 565-0871 (Japan); Takaki, N.; Yamaguchi, A.; Watanabe, H. [Tokai Univ., 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa, 259-1292 (Japan); Unesaki, H. [Kyoto Univ. Research Reactor Inst., Asahiro-nishi, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan)

    2012-07-01

    Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

  18. A Pin Power Reconstruction Method for CANDU Reactor Cores Based on Coarse-Mesh Finite Difference Calculations

    SciTech Connect

    Lee, Hyung-Seok [Chosun University (Korea, Republic of); Yang, Won Sik [Chosun University (Korea, Republic of); Na, Man Gyun [Chosun University (Korea, Republic of); Choi, Hangbok [Korea Atomic Energy Research Institute (Korea, Republic of)

    2000-04-15

    A reconstruction method has been developed for recovering pin powers from Canada deuterium uranium (CANDU) reactor core calculations performed with a coarse-mesh finite difference diffusion approximation and single-assembly lattice calculations. The homogeneous intranodal distributions of group fluxes are efficiently computed using polynomial shapes constrained to satisfy the nodal information approximated from the node-average fluxes. The group fluxes of individual fuel pins in a heterogeneous fuel bundle are determined using these homogeneous intranodal flux distributions and the form functions obtained from the single-assembly lattice calculations. The pin powers are obtained using these pin fluxes and the pin power cross sections generated by the single-assembly lattice calculation. The accuracy of the reconstruction schemes has been estimated by performing benchmark calculations for partial core representation of a natural uranium CANDU reactor. The results indicate that the reconstruction schemes are quite accurate, yielding maximum pin power errors of less than {approx}3%. The main contribution to the reconstruction error is made by the errors in the node-average fluxes obtained from the coarse-mesh finite difference diffusion calculation; the errors due to the reconstruction schemes are <1%.

  19. Calculation of evanescent field interaction with metallic nanoparticles immobilized on the air holes of solid-core photonic crystal fiber

    NASA Astrophysics Data System (ADS)

    Raspopin, Alexander; Cui, Hong-Liang; Du, Henry

    2005-11-01

    We report the calculation of the attenuation coefficient of a probing optical mode due to interaction with metallic nanoparticles randomly distributed in the air holes of a solid core photonic crystal fiber (PCF) for SERS-based sensing and detection. The approach employed is an approximation of the solid core PCF with conventional curricular fiber considering a similar total internal reflection mechanism for mode propagation and almost complete concentration of the mode flux in the core of PCF. Losses due to the absorption and radiative scattering of electromagnetic energy by nanoparticles are examined. The analysis demonstrates a critical dependence of the absorption losses approaching the resonant localized surface plasmon's excitation and very fast rise of radiative losses with the increase of nanoparticle size. The physics of proper integral Raman spectroscopy is also discussed.

  20. Detailed microscopic calculation of stellar electron and positron capture rates on $^{24}$Mg for O+Ne+Mg core simulations

    E-print Network

    Nabi, Jameel-Un

    2014-01-01

    Few white dwarfs, located in binary systems, may acquire sufficiently high mass accretion rates resulting in the burning of carbon and oxygen under nondegenerate conditions forming a O+Ne+Mg core. These O+Ne+Mg cores are gravitationally less bound than more massive progenitor stars and can release more energy due to the nuclear burning. They are also amongst the probable candidates for low entropy r-process sites. Recent observations of subluminous Type II-P supernovae (e.g., 2005cs, 2003gd, 1999br, 1997D) were able to rekindle the interest in 8 -- 10 M$_{\\odot}$ which develop O+Ne+Mg cores. Microscopic calculations of capture rates on $^{24}$Mg, which may contribute significantly to the collapse of O+Ne+Mg cores, using shell model and proton-neutron quasiparticle random phase approximation (pn-QRPA) theory, were performed earlier and comparisons made. Simulators, however, may require these capture rates on a fine scale. For the first time a detailed microscopic calculation of the electron and positron captur...

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

    SciTech Connect

    Killeen, J.C. [International Atomic Energy Agency, Wagramerstrasse 5, PO Box 100, A-1400 Vienna (Austria); Turnbull, J.A. [Cherry-Lyn, Tockington, South Glos (United Kingdom); Sartori, E. [OECD/NEA, 12 Bd des Iles, 92130 Issy-les-Moulineaux (France)

    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 internal modelling. (authors)

  2. Dependence of transuranic content in spent fuel on fuel burnup

    E-print Network

    Reese, Drew A. (Drew Amelia)

    2007-01-01

    As the increasing demand for nuclear energy results in larger spent fuel volume, implementation of longer fuel cycles incorporating higher burnup are becoming common. Understanding the effect of higher burnup on the spent ...

  3. Transmission electron microscopy and ab initio calculations to relate interfacial intermixing and the magnetism of core/shell nanoparticles

    NASA Astrophysics Data System (ADS)

    Chi, C.-C.; Hsiao, C.-H.; Skoropata, E.; van Lierop, J.; Ouyang, Chuenhou Hao

    2015-05-01

    Significant efforts towards understanding bi-magnetic core-shell nanoparticles are underway currently as they provide a pathway towards properties unavailable with single-phased systems. Recently, we have demonstrated that the magnetism of ?-Fe2O3/CoO core-shell nanoparticles, in particular, at high temperatures, originates essentially from an interfacial doped iron-oxide layer that is formed by the migration of Co2+ from the CoO shell into the surface layers of the ?-Fe2O3 core [Skoropata et al., Phys. Rev. B 89, 024410 (2014)]. To examine directly the nature of the intermixed layer, we have used high-resolution transmission electron microscopy (HRTEM) and first-principles calculations to examine the impact of the core-shell intermixing at the atomic level. By analyzing the HRTEM images and energy dispersive spectra, the level and nature of intermixing was confirmed, mainly as doping of Co into the octahedral site vacancies of ?-Fe2O3. The average Co doping depths for different processing temperatures (150 °C and 235 °C) were 0.56 nm and 0.78 nm (determined to within 5% through simulation), respectively, establishing that the amount of core-shell intermixing can be altered purposefully with an appropriate change in synthesis conditions. Through first-principles calculations, we find that the intermixing phase of ?-Fe2O3 with Co doping is ferromagnetic, with even higher magnetization as compared to that of pure ?-Fe2O3. In addition, we show that Co doping into different octahedral sites can cause different magnetizations. This was reflected in a change in overall nanoparticle magnetization, where we observed a 25% reduction in magnetization for the 235 °C versus the 150 °C sample, despite a thicker intermixed layer.

  4. 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. (Los Alamos National Laboratory, Los Alamos, NM); 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.

  5. 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.

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

    SciTech Connect

    Cabellos, O. [Universidad Politecnica de Madrid, Dpto. Ingenieria Nuclear, Madrid (Spain); Sanz, J.; Rodriguez, A. [Univ. National Educacion a Distancia, Dpto. Ingenieria Energetica, Madrid (Spain); Gonzalez, E.; Embid, M.; Alvarez, F. [CIEMAT, Madrid (Spain); Reyes, S. [Lawrence Livermore National Laboratory, Livermore CA (United States)

    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.

  7. Component mode synthesis methods applied to 3D heterogeneous core calculations, using the mixed dual finite element solver MINOS

    SciTech Connect

    Guerin, P.; Baudron, A. M.; Lautard, J. J. [Commissariat a l'Energie Atomique, DEN/DANS/DM2S/SERMA/LENR, CEA Saclay, 91191 Gif sur Yvette (France)

    2006-07-01

    This paper describes a new technique for determining the pin power in heterogeneous core calculations. It is based on a domain decomposition with overlapping sub-domains and a component mode synthesis technique for the global flux determination. Local basis functions are used to span a discrete space that allows fundamental global mode approximation through a Galerkin technique. Two approaches are given to obtain these local basis functions: in the first one (Component Mode Synthesis method), the first few spatial eigenfunctions are computed on each sub-domain, using periodic boundary conditions. In the second one (Factorized Component Mode Synthesis method), only the fundamental mode is computed, and we use a factorization principle for the flux in order to replace the higher order Eigenmodes. These different local spatial functions are extended to the global domain by defining them as zero outside the sub-domain. These methods are well-fitted for heterogeneous core calculations because the spatial interface modes are taken into account in the domain decomposition. Although these methods could be applied to higher order angular approximations - particularly easily to a SPN approximation - the numerical results we provide are obtained using a diffusion model. We show the methods' accuracy for reactor cores loaded with UOX and MOX assemblies, for which standard reconstruction techniques are known to perform poorly. Furthermore, we show that our methods are highly and easily parallelizable. (authors)

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

    NASA Astrophysics Data System (ADS)

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

    2011-12-01

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

  9. 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.

  10. 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 ?-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 perturbations and then transformed through the Physical Programming (PP) paradigm to build an aggregate objective function without iteratively determining weights. Finally, the design is optimized with standard gradient-based methods. Through the power of ACE and the transparency of PP, the optimization system allows users to locate designs that best suit their multiobjective preferences with ease.

  11. AB initio free energy calculations of the solubility of silica in metallic hydrogen and application to giant planet cores

    SciTech Connect

    González-Cataldo, F. [Grupo de NanoMateriales, Departamento de Física, Facultad de Ciencias, Universidad de Chile, Casilla 653, Santiago (Chile); Wilson, Hugh F.; Militzer, B., E-mail: fgonzalez@lpmd.cl [Department of Earth and Planetary Science, University of California Berkeley, Berkeley, CA 94720 (United States)

    2014-05-20

    By combining density functional molecular dynamics simulations with a thermodynamic integration technique, we determine the free energy of metallic hydrogen and silica, SiO{sub 2}, at megabar pressures and thousands of degrees Kelvin. Our ab initio solubility calculations show that silica dissolves into fluid hydrogen above 5000 K for pressures from 10 and 40 Mbars, which has implications for the evolution of rocky cores in giant gas planets like Jupiter, Saturn, and a substantial fraction of known extrasolar planets. Our findings underline the necessity of considering the erosion and redistribution of core materials in giant planet evolution models, but they also demonstrate that hot metallic hydrogen is a good solvent at megabar pressures, which has implications for high-pressure experiments.

  12. Iron at Earth's core conditions from first principles calculations Dario Alf`e

    E-print Network

    Alfè, Dario

    , and I will describe how using the molecular dynamics technique, coupled with ab-initio calculations 2.2 Elastic constants will help us to interpret and hopefully predict the behaviour of the dynamical processes that occour inside

  13. Three-Dimensional Full Core Power Calculations for Pressurized Water Reactors

    SciTech Connect

    Evans, Thomas M [ORNL] [ORNL; Davidson, Gregory G [ORNL] [ORNL; Slaybaugh, Rachel N [ORNL] [ORNL

    2010-01-01

    We have implemented a new multilevel parallel decomposition in the Denovo discrete ordinates radiation transport code. In concert with Krylov subspace iterative solvers, the multilevel decomposition allows concurrency over energy in addition to space-angle. The original space-angle partitioning in Denovo placed an eective limit on the scalability of the transport solver that was highly dependent on the problem size. The added phase-space concurrency combined with the high-performance Krylov solvers has enabled weak scaling to 100K cores on the Jaguar XT5 supercomputer. Furthermore, the multilevel decomposition provides enough concurrency to scale to exascale computing and beyond.

  14. Numerical calculation of the radiation exposure from galactic cosmic rays at aviation altitudes with the PANDOCA core model

    NASA Astrophysics Data System (ADS)

    Matthiä, Daniel; Meier, Matthias M.; Reitz, Günther

    2014-03-01

    The increased radiation exposure at aviation altitudes is of public interest as well as of legal relevance in many countries. The dose rates that are elevated compared to sea level are mainly caused by galactic cosmic ray particles interacting with the atmosphere and producing a complex radiation field at aviation altitudes. The intensity and composition of this radiation field mainly depend on altitude, geomagnetic shielding, and primary particle intensity. In this work, we present a model based on Monte Carlo simulations, which retrospectively estimates secondary particle fluence as well as ambient dose equivalent rates and effective dose rates at any point in the atmosphere. This model will be used as the physical core in the Professional Aviation Dose Calculator (PANDOCA) software developed by the German Aerospace Center (Deutsches Zentrum für Luft- und Raumfahrt) for the calculation of route doses in aviation. The calculations are based on galactic cosmic ray spectra taking into account primary nuclei from hydrogen to iron by direct transport calculations of hydrogen and helium nuclei and approximating heavier nuclei by the number of protons equaling the corresponding atomic number. A comparison to experimental data recorded on several flights with a tissue equivalent proportional counter shows a very good agreement between model calculations and measurements.

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

    SciTech Connect

    Suyama, Kenya [Japan Atomic Energy Research Institute (Japan); Mochizuki, Hiroki [Japan Atomic Energy Research Institute (Japan); Kiyosumi, Takehide [Japan Research Institute, Ltd. (Japan)

    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.

  16. Three Dimensional Analysis of 3-Loop PWR RCCA Ejection Accident for High Burnup

    SciTech Connect

    Marciulescu, Cristian; Sung, Yixing; Beard, Charles L. [Westinghouse Electric Company, LLC (United States)

    2006-07-01

    The Rod Control Cluster Assembly (RCCA) ejection accident is a Condition IV design basis reactivity insertion event for Pressurized Water Reactors (PWR). The event is historically analyzed using a one-dimensional (1D) neutron kinetic code to meet the current licensing criteria for fuel rod burnup to 62,000 MWD/MTU. The Westinghouse USNRC-approved three-dimensional (3D) analysis methodology is based on the neutron kinetics version of the ANC code (SPNOVA) coupled with Westinghouse's version of the EPRI core thermal-hydraulic code VIPRE-01. The 3D methodology provides a more realistic yet conservative analysis approach to meet anticipated reduction in the licensing fuel enthalpy rise limit for high burnup fuel. A rod ejection analysis using the 3D methodology was recently performed for a Westinghouse 3-loop PWR at an up-rated core power of 3151 MWt with reload cores that allow large flexibility in assembly shuffling and a fuel hot rod burnup to 75,000 MWD/MTU. The analysis considered high enrichment fuel assemblies at the control rod locations as well as bounding rodded depletions in the end of life, zero power and full power conditions. The analysis results demonstrated that the peak fuel enthalpy rise is less than 100 cal/g for the transient initiated at the hot zero power condition. The maximum fuel enthalpy is less than 200 cal/g for the transient initiated from the full power condition. (authors)

  17. The effect of nickel on the properties of iron at the conditions of Earth's inner core: Ab initio calculations of seismic wave velocities of

    E-print Network

    Vocadlo, Lidunka

    The effect of nickel on the properties of iron at the conditions of Earth's inner core: Ab initio calculations of seismic wave velocities of Fe­Ni alloys Benjami´ Martorell n , John Brodholt, Ian G. Wood­Ni alloy high pressure high temperature inner core elastic properties compressional and shear wave

  18. 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, they do demonstrate that the effect of BPRs is generally well behaved and that independent codes and cross-section libraries predict similar results. The report concludes with a discussion of the issues for consideration and recommendations for inclusion of SNF assemblies exposed to BPRs in criticality safety analyses using burnup credit for dry cask storage and transport.

  19. Comparison of TRAC-PD2 calculations to emergency-core-coolant bypass data

    SciTech Connect

    Meier, J.K.

    1982-01-01

    The Los Alamos Transient Reactor Analysis Code, TRAC-PD2, is a multidimensional thermal-hydraulic computer code that can simulate loss-of-coolant accidents (LOCAs) in pressurized-water reactors (PWRs). As part of our independent assessment of this code, we compared TRAC calculated results with Battelle Columbus Laboratories (BCL) test data. We based our comparisons on the BCL downcomer countercurrent flow test that used a 2/15-scale model of a coreless PWR vessel. In general, TRAC tended to underpredict the steam flow necessary to prevent significant flooding of the downcomer. This underprediction is insignificant in low-subcooling cases; however, in high-subcooling cases, the steam flows for similar degrees of water penetration were 30 to 35% less in the TRAC predictions than in the test data. Thus, we infer that the calculations for the momentum transfer between the two phases were adequate, but that the interfacial heat-transfer correlations could use improvement. 14 figures.

  20. First-principles calculations of properties of orthorhombic iron carbide Fe7C3 at the Earth's core conditions

    NASA Astrophysics Data System (ADS)

    Raza, Zamaan; Shulumba, Nina; Caffrey, Nuala M.; Dubrovinsky, Leonid; Abrikosov, Igor A.

    2015-06-01

    A recently discovered phase of orthorhombic iron carbide o-Fe7C3 [Prescher et al., Nat. Geosci. 8, 220 (2015), 10.1038/ngeo2370] is assessed as a potentially important phase for interpretation of the properties of the Earth's core. In this paper, we carry out first-principles calculations on o-Fe7C3 , finding properties to be in broad agreement with recent experiments, including a high Poisson's ratio (0.38). Our enthalpy calculations suggest that o-Fe7C3 is more stable than Eckstrom-Adcock hexagonal iron carbide (h-Fe7C3 ) below approximately 100 GPa. However, at 150 GPa, the two phases are essentially degenerate in terms of Gibbs free energy, and further increasing the pressure towards Earth's core conditions stabilizes h-Fe7C3 with respect to the orthorhombic phase. Increasing the temperature tends to stabilize the hexagonal phase at 360 GPa, but this trend may change beyond the limit of the quasiharmonic approximation.

  1. Extended burnup demonstration reactor fuel program. Semiannual progress report, October 1979-March 1980. Report XN-NF-80-26

    SciTech Connect

    Woods, K.N.; Van Swam, L.F.

    1980-12-31

    The first of three scheduled poolside fuel examinations at the Oyster Creek reactor conducted during February/March 1980, was directed at one of the four symmetrically loaded ENC 8 x 8 lead assemblies that had achieved a burnup of approx. 25,000 MWd/MTU. Forty-five of the fuel rods in assembly UD3-109 were removed and examined. In general, the individual fuel rods were in excellent condition. The average fuel rod diameter continued to decrease during the last cycle was assembly burnup increased from 19,500 to 25,700 MWd/MTU. The creepdown since the beginning of life (BOL) in the center of the fuel rods is about 0.003 in. The fuel rods bore no indication of cladding ridging. Fuel rod growth continued at a linear rate of about 0.02% per GWd/MTU burnup since BOL. Preliminary eddy current test data showed that the cladding was free of significant defects. Visual examination of fuel rods, spacers, and tie plate revealed no unusual conditions. An average assembly burnup of 31,000 MWd/MTU is projected by the end of the next reactor cycle. Calculated plastic strain and internal rod pressure were shown to be within the design limit at that burnup. Licensing approval was obtained for the first extended burnup cycle. A cryogenically cooled germanium gamma scanning system for use at poolside was designed.

  2. First-principles calculation of electronic stopping contributions from core electrons and off-channeling

    NASA Astrophysics Data System (ADS)

    Correa, Alfredo; Schleife, Andre; Kanai, Yosuke

    2014-03-01

    In order to understand the interaction of projectile atoms with targets under particle radiation in materials, e.g. in space applications or nuclear reactors, it is critical to investigate electronic and ionic contributions to stopping power. The goal of such efforts is detailed understanding of radiation damages as well as fundamental effects such as ion-electron interaction. While ionic stopping has been successfully modeled by molecular dynamics in the past, only recently a computational framework came within reach that is capable of accurately describing electronic stopping from first principles. Using our large-scale implementation of real-time time-dependent density functional theory in non-adiabatic Ehrenfest molecular dynamics, we are able to gain deep insight into electronic stopping for systems with hundreds of atoms and thousands of electrons, taking into account their quantum-mechanical electron-electron interaction. We discuss distinct contributions of valence and core electrons of aluminum target atoms to electronic stopping, and we study their importance for different projectile (hydrogen and helium atoms) velocities. There is striking influence of the stopping geometry especially for fast projectiles, and we find excellent agreement with experiment. Prepared by LLNL under Contract DE-AC52-07NA27344.

  3. Analysis and comparison of CVS-ADC approaches up to third order for the calculation of core-excited states.

    PubMed

    Wenzel, Jan; Holzer, Andre; Wormit, Michael; Dreuw, Andreas

    2015-06-01

    The extended second order algebraic-diagrammatic construction (ADC(2)-x) scheme for the polarization operator in combination with core-valence separation (CVS) approximation is well known to be a powerful quantum chemical method for the calculation of core-excited states and the description of X-ray absorption spectra. For the first time, the implementation and results of the third order approach CVS-ADC(3) are reported. Therefore, the CVS approximation has been applied to the ADC(3) working equations and the resulting terms have been implemented efficiently in the adcman program. By treating the ? and ? spins separately from each other, the unrestricted variant CVS-UADC(3) for the treatment of open-shell systems has been implemented as well. The performance and accuracy of the CVS-ADC(3) method are demonstrated with respect to a set of small and middle-sized organic molecules. Therefore, the results obtained at the CVS-ADC(3) level are compared with CVS-ADC(2)-x values as well as experimental data by calculating complete basis set limits. The influence of basis sets is further investigated by employing a large set of different basis sets. Besides the accuracy of core-excitation energies and oscillator strengths, the importance of cartesian basis functions and the treatment of orbital relaxation effects are analyzed in this work as well as computational timings. It turns out that at the CVS-ADC(3) level, the results are not further improved compared to CVS-ADC(2)-x and experimental data, because the fortuitous error compensation inherent in the CVS-ADC(2)-x approach is broken. While CVS-ADC(3) overestimates the core excitation energies on average by 0.61% ± 0.31%, CVS-ADC(2)-x provides an averaged underestimation of -0.22% ± 0.12%. Eventually, the best agreement with experiments can be achieved using the CVS-ADC(2)-x method in combination with a diffuse cartesian basis set at least at the triple-? level. PMID:26049476

  4. Analysis and comparison of CVS-ADC approaches up to third order for the calculation of core-excited states

    NASA Astrophysics Data System (ADS)

    Wenzel, Jan; Holzer, Andre; Wormit, Michael; Dreuw, Andreas

    2015-06-01

    The extended second order algebraic-diagrammatic construction (ADC(2)-x) scheme for the polarization operator in combination with core-valence separation (CVS) approximation is well known to be a powerful quantum chemical method for the calculation of core-excited states and the description of X-ray absorption spectra. For the first time, the implementation and results of the third order approach CVS-ADC(3) are reported. Therefore, the CVS approximation has been applied to the ADC(3) working equations and the resulting terms have been implemented efficiently in the adcman program. By treating the ? and ? spins separately from each other, the unrestricted variant CVS-UADC(3) for the treatment of open-shell systems has been implemented as well. The performance and accuracy of the CVS-ADC(3) method are demonstrated with respect to a set of small and middle-sized organic molecules. Therefore, the results obtained at the CVS-ADC(3) level are compared with CVS-ADC(2)-x values as well as experimental data by calculating complete basis set limits. The influence of basis sets is further investigated by employing a large set of different basis sets. Besides the accuracy of core-excitation energies and oscillator strengths, the importance of cartesian basis functions and the treatment of orbital relaxation effects are analyzed in this work as well as computational timings. It turns out that at the CVS-ADC(3) level, the results are not further improved compared to CVS-ADC(2)-x and experimental data, because the fortuitous error compensation inherent in the CVS-ADC(2)-x approach is broken. While CVS-ADC(3) overestimates the core excitation energies on average by 0.61% ± 0.31%, CVS-ADC(2)-x provides an averaged underestimation of -0.22% ± 0.12%. Eventually, the best agreement with experiments can be achieved using the CVS-ADC(2)-x method in combination with a diffuse cartesian basis set at least at the triple-? level.

  5. Burnup calculations for the HOMER15 and SAFE300 reactors

    Microsoft Academic Search

    Benjamin W. Amiri; David I. Poston

    2002-01-01

    The Heatpipe Power System (HPS) is a near-term low-cost space fission power system. As the U-235 fuel of the HPS is burned, higher actinides and fission products will be produced. This will cause changes in system reactivity, radioactivity, and decay power. One potential concern is that gaseous fission products may exert excessive pressure on the fuel pin cladding. To evaluate

  6. Evaluation of accuracy of calculations of VVER-1000 core states with incomplete covering of fuel by the absorber

    SciTech Connect

    Tikhomirov, A. V.; Ponomarenko, G. L. [OKB GIDROPRESS, Podolsk (Russian Federation)

    2012-07-01

    An additional verification of bundled software (BS) SAPFIR-95 and amp;RC [1] and code KORSAR/GP [2] was performed. Both software products were developed in A.P. Alexandrov NITI and certified by ROSTEKHNADZOR of RF for numeric simulation of stationary, transitional and emergency conditions of VVER reactors. A benchmark model for neutronics calculations was created within the limits of this work. The cold subcritical state of VVER - 1000 reactor stationary fuelling was simulated on the basis of FA with an increased height of the fuel column (TVS-2M) considering detailed presentation of radial and front neutron reflectors. A case of passing of pure condensate slug through the core in initially deep subcritical state during start of the first RCP set after refueling was considered as an examined condition of reactor operation. A relatively small size of the slug, its spatial position near the reflectors (lower and lateral), as well as failure of the inserted control rods of the control and protection system (CPS CR) to reach the lower limit of the fuel column stipulate for methodical complexity of a correct calculation of the neutron multiplication constant (K{sub eff}) using engineering codes. Code RC was used as a test program in the process of reactor calculated 3-D modeling. Code MCNP5 [3] was used as the precision program, which solves the equation of neutrons transfer by Monte-Carlo method and which was developed in the US (Los-Alamos). As a result of comparative calculations dependency of K{sub eff} on two parameters was evaluated - boron acid concentration (Cb) and CPS CR position. Reactivity effect was evaluated, which is implemented as a result of failure of all CPS control rods to reach the lower fuel limit calculated using the engineering codes mentioned above. (authors)

  7. Burnup dependence of melting temperature of FBR mixed oxide fuels irradiated to high burnup

    NASA Astrophysics Data System (ADS)

    Hirosawa, Takashi; Sato, Isamu

    2011-11-01

    The melting temperatures of FBR MOX fuels with Pu content of 28-30 wt.% irradiated to from 22.5 to 112.5 MWd kg -1 were measured using a rhenium inner capsule to hold the specimens. The rhenium inner capsule could prevent chemical reactions between fuels and tungsten materials which decrease the melting temperature. The melting temperatures were about 30 K higher than the previous data using tungsten capsules. The melting temperature decreases in a linear manner with burnup due to solid solution of fission products in fuels. However, the slopes of the lines plotting melting temperature versus burnup are almost similar to the previous data.

  8. 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.

  9. Hybrid density functional-molecular mechanics calculations for core-electron binding energies of glycine in water solution.

    PubMed

    Niskanen, Johannes; Arul Murugan, N; Rinkevicius, Zilvinas; Vahtras, Olav; Li, Cui; Monti, Susanna; Carravetta, Vincenzo; Agren, Hans

    2013-01-01

    We report hybrid density functional theory-molecular mechanics (DFT/MM) calculations performed for glycine in water solution at different pH values. In this paper, we discuss several aspects of the quantum mechanics-molecular mechanics (QM/MM) simulations where the dynamics and spectral binding energy shifts are computed sequentially, and where the latter are evaluated over a set of configurations generated by molecular or Car-Parrinello dynamics simulations. In the used model, core ionization takes place in glycine as a quantum mechanical (QM) system modeled with DFT, and the solution is described with expedient force fields in a large molecular mechanical (MM) volume of water molecules. The contribution to the core electronic binding energy from all interactions within and between the two (DFT and MM) parts is accounted for, except charge transfer and dispersion. While the obtained results were found to be in qualitative agreement with experiment, their precision must be qualified with respect to the problem of counter ions, charge transfer and optimal division of QM and MM parts of the system. Results are compared to those of a recent study [Ottoson et al., J. Am. Chem. Soc., 2011, 133, 3120]. PMID:23160171

  10. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  11. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    PubMed Central

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

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

    NASA Astrophysics Data System (ADS)

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

    2006-01-01

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

  13. 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.

  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. DANDE: a linked code system for core neutronics/depletion analysis

    SciTech Connect

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1986-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs.

  16. A Tight Lattice, Epithermal Core Design for the Integral PWR

    SciTech Connect

    Saccheri, J.G.B. [Brookhaven National Laboratory, Nuclear Science and Technology Division Bldg 475, Upton, New York 11973-5000 (United States); Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge, Bldg. 24-205 MA 02139-4307 (United States)

    2004-07-01

    An 8-year core design for an epithermal, water-cooled reactor has been developed based upon assessments of nuclear reactor physics, thermal-hydraulics and economics. An integral vessel configuration is adopted and self-supporting wire-wrap fuel is employed for the tight lattice of the epithermal core. A streaming path is incorporated in each assembly to ensure a negative void coefficient. A whole-core MCNP simulation of the tight core shows a negative void coefficient for any burnup with positive K{sub EFF}. The VIPRE{sup TM} code has been used to calculate the critical heat flux (CHF) by means of an appropriate wire-wrap CHF correlation, specifically introduced in the source code. Economically, the high fuel enrichment (14% w/o {sup 235}U) and the very long core life (8 ys) lead to high lifetime-levelized unit fuel cycle cost (in mills/kWhre). However, both operation and maintenance and capital-related expenditures strongly benefited from the higher electric output per unit volume, which yielded quite small lifetime-levelized unit capital and operation and maintenance costs for the overall plant. Financing costs are included and an estimate is provided for the total lifetime-levelized unit cost of the epithermal core, which is about 20% lower than that of a more open lattice thermal spectrum core fitting into the same core envelope and with 4-year lifetime. (authors)

  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. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    SciTech Connect

    Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2011-07-01

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

  19. Depletion calculations for the McClellan Nuclear Radiation Center.

    SciTech Connect

    Klann, R. T.; Newell, D. L.

    1997-12-08

    Depletion calculations have been performed for the McClellan reactor history from January 1990 through August 1996. A database has been generated for continuing use by operations personnel which contains the isotopic inventory for all fuel elements and fuel-followed control rods maintained at McClellan. The calculations are based on the three-dimensional diffusion theory code REBUS-3 which is available through the Radiation Safety Information Computational Center (RSICC). Burnup-dependent cross-sections were developed at zero power temperatures and full power temperatures using the WIMS code (also available through RSICC). WIMS is based on discretized transport theory to calculate the neutron flux as a function of energy and position in a one-dimensional cell. Based on the initial depletion calculations, a method was developed to allow operations personnel to perform depletion calculations and update the database with a minimal amount of effort. Depletion estimates and calculations can be performed by simply entering the core loading configuration, the position of the control rods at the start and end of cycle, the reactor power level, the duration of the reactor cycle, and the time since the last reactor cycle. The depletion and buildup of isotopes of interest (heavy metal isotopes, erbium isotopes, and fission product poisons) are calculated for all fuel elements and fuel-followed control rods in the MNRC inventory. The reactivity loss from burnup and buildup of fission product poisons and the peak xenon buildup after shutdown are also calculated. The reactivity loss from going from cold zero power to hot full power can also be calculated by using the temperature-dependent, burnup-dependent cross-sections. By calculating all of these reactivity effects, operations personnel are able to estimate the total excess reactivity necessary to run the reactor for the given cycle. This method has also been used to estimate the worth of individual control rods. Using this approach, fuel management and core loading can be optimized such that each individual fuel element and fuel-followed control rod is used to its full potential before being replaced with fresh fuel. This fuel management strategy allows a significant cost saving to MNRC by reducing fuel replacement costs and maximizing the usefulness of each element in the inventory.

  20. Building neutron cross-section dependencies for few-group reactor calculations using stepwise regression

    Microsoft Academic Search

    Vyacheslav G. Zimin; Andrey A. Semenov

    2005-01-01

    Approximation of few-group neutron cross-sections by functions of burnup and thermal-hydraulics parameters of a fuel cell is considered. The cross-section is written as a sum of two terms: the base cross-section, which depends only on burnup and is computed under the nominal reactor core conditions, and the deviation, which depends on burnup and thermal-hydraulics variables of the cell. A one-dimensional

  1. Transverse buckling effects on solitary burn-up waves

    Microsoft Academic Search

    Xue-Nong Chen; Werner Maschek

    2005-01-01

    A three-dimensional one-group diffusion model with explicit effects of burnup and feedback is studied for a so-called “candle reactor”. By a perturbation method the problem is reduced to a one-dimensional one, for which a solitary wave solution was obtained by van Dam (2000) [Self-stabilizing criticality waves. Annals of Nuclear Energy 27, 1505]. Therefore, such a travelling burn-up wave exists as

  2. Calculated Coupling Efficiency Between an Elliptical-Core Optical Fiber and a Silicon Oxynitride Rib Waveguide [Corrected Copy

    NASA Technical Reports Server (NTRS)

    Tuma, Margaret L.; Beheim, Glenn

    1995-01-01

    The effective-index method and Marcatili's technique were utilized independently to calculate the electric field profile of a rib channel waveguide. Using the electric field profile calculated from each method, the theoretical coupling efficiency between a single-mode optical fiber and a rib waveguide was calculated using the overlap integral. Perfect alignment was assumed and the coupling efficiency calculated. The coupling efficiency calculation was then repeated for a range of transverse offsets.

  3. Use of Axially Graded Burnable Boron for Hot-Spot Temperature Reduction in a Pressurized Water Reactor Core

    SciTech Connect

    Segev, M.; Galperin, A.; Schwageraus, E. [Ben Gurion University of the Negev (Israel)

    2000-07-15

    Shortly after the loading of a pressurized water reactor (PWR) core, the axial power distribution in fresh fuel has already attained the characteristic, almost flat shape, typical of burned fuel. At beginning of cycle (BOC), however, the axial distribution is centrally peaked. In assemblies hosting uniform burnable boron rods, this BOC peaking is even more pronounced. A reduction in the axial peaking is today often achieved by shortening the burnable boron rods by some 30 cm at each edge.It is shown that a two-zone grading of the boron rod leads, in a representative PWR cycle, to a reduction of the hot-spot temperature of {approx}70 deg. C, compared with the nongraded case. However, with a proper three-zone grading of the boron rod, an additional 20 deg. C may be cut off the hot-spot temperature. Further, with a slightly skewed application of this three-zone grading, an additional 50 deg. C may be cut off.The representative PWR cycle studied was cycle 11 of the Indian Point 2 station, with a simplification in the number of fuel types and in the burnup distribution. The analysis was based on a complete three-dimensional burnup calculation. The code system was ELCOS, with BOXER as an assembly code for the generation of burnup-dependent cross sections and SILWER as a three-dimensional core code with thermal-hydraulic feedback.

  4. Analysis of MNSR core composition changes using the codes WIMSD-4 and CITATION.

    PubMed

    Haj Hassan, H; Ghazi, N; Hainoun, A

    2008-10-01

    The codes WIMSD/4 and BORGES--part of the MTR-PC code package--have been applied to prepare the microscopic cross-section library for the main elements of miniature neutron source reactor (MNSR) core for six neutron energy groups. The generated library has been utilized by the 3D code CITATION to perform the calculation of fuel burn-up including the identification of main fission products and their impacts on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products. The burn-up results have indicated that the core life-time of MNSR is being mainly estimated by Sm(149) followed by Gd(157) and Cd(113). The accumulation of these fission products during 100 continuous operation days caused a reduction of about 4.3 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 3.5 mk, which relates to the whole discontinuous operation period of the reactor since its start up to now. The calculation procedure simulates the sporadic operation with an equivalent continuous operation period. This approximation is valid for the long-lived fission products that mainly dictate the core life-time. However, it is an overestimation for the concentration of short-lived radioactive products like Xe(135). PMID:18547812

  5. 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. [Commissariat a l'Energie Atomique et aux Energies Alternatives CEA, Service d'Etude des Reacteurs et de Mathematiques Appliquees, DEN/DANS/DM2S/SERMA/LTSD, F91191 Gif-sur-Yvette cedex (France)

    2013-07-01

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

  6. 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.

  7. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    NASA Astrophysics Data System (ADS)

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen

    2005-05-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  8. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    SciTech Connect

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy [Institute for Nuclear Research, Prospekt Nauky 47, Kyiv, 03680 (Ukraine); Binney, Stephen [Oregon State University, Corvallis, OR 97331-5902 (United States)

    2005-05-24

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  9. Ground-state properties of alkali dimers and their cations (including the elements Li, Na, and K) from ab initio calculations with effective core polarization potentials

    Microsoft Academic Search

    Wolfgang Müller; Wilfried Meyer

    1984-01-01

    Extensive all-electron SCF and valence CI calculations are presented for alkali dimer systems with consideration of intershell correlation effects by use of an effective core polarization potential (CPP), which contains only a single adjustable atomic parameter. High accuracy is obtained for the ground-state spectroscopic constants of the studied molecules. The maximum deviations from accurate experimental data are as follows: 1%

  10. PWR cores with silicon carbide cladding

    SciTech Connect

    Dobisesky, J. P.; Carpenter, D.; Pilat, E.; Kazimi, M. S. [Center for Advanced Nuclear Energy Systems, Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, 77 Massachusetts Avenue 24-215, Cambridge, MA 02139-4307 (United States)

    2012-07-01

    The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

  11. PAMELA: An open-source software package for calculating nonlocal exact exchange effects on electron gases in core-shell nanowires

    NASA Astrophysics Data System (ADS)

    Long, Andrew W.; Wong, Bryan M.

    2012-09-01

    We present a new pseudospectral approach for incorporating many-body, nonlocal exact exchange interactions to understand the formation of electron gases in core-shell nanowires. Our approach is efficiently implemented in the open-source software package PAMELA (Pseudospectral Analysis Method with Exchange & Local Approximations) that can calculate electronic energies, densities, wavefunctions, and band-bending diagrams within a self-consistent Schrödinger-Poisson formalism. The implementation of both local and nonlocal electronic effects using pseudospectral methods is key to PAMELA's efficiency, resulting in significantly reduced computational effort compared to finite-element methods. In contrast to the new nonlocal exchange formalism implemented in this work, we find that the simple, conventional Schrödinger-Poisson approaches commonly used in the literature (1) considerably overestimate the number of occupied electron levels, (2) overdelocalize electrons in nanowires, and (3) significantly underestimate the relative energy separation between electronic subbands. In addition, we perform several calculations in the high-doping regime that show a critical tunneling depth exists in these nanosystems where tunneling from the core-shell interface to the nanowire edge becomes the dominant mechanism of electron gas formation. Finally, in order to present a general-purpose set of tools that both experimentalists and theorists can easily use to predict electron gas formation in core-shell nanowires, we document and provide our efficient and user-friendly PAMELA source code that is freely available at http://alum.mit.edu/www/usagi.

  12. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    SciTech Connect

    Pecchia, M.; D'Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  13. First 3-D calculation of core disruptive accident in a large-scale sodium-cooled fast reactor

    Microsoft Academic Search

    Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita; Werner Maschek

    2009-01-01

    The SIMMER-IV computer code is a three-dimensional fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. The present study has attempted the first application of SIMMER-IV to a core disruptive accident in a large-scale sodium-cooled fast reactor. A principal point of this study was to investigate reactivity effects with fuel relocation under three-dimensional

  14. Comments on fission-gas release from fuel at high burnup in Vol. 19, No. 6. [Water cooled reactors

    Microsoft Academic Search

    H. Ocken; J. T. A. Roberts

    1979-01-01

    Meyer, Beyer, and Voglewede have proposed that an enhancement factor be applied to existing vendor models when fission-gas release (FGR) at burnups greater than 20,000 MWd\\/metric ton is calculated for licensing purposes. This enhancement factor is derived from FGR data obtained from liquid-metal-cooled fast breeder reactor (LMFBR) fuel. The analysis assumes that the intrinsic source of the high FGR measured

  15. Isotopic biases for actinide-only burnup credit

    SciTech Connect

    Rahimi, M.; Lancaster, D.; Hoeffer, B.; Nichols, M.

    1997-04-01

    The primary purpose of this paper is to present the new methodology for establishing bias and uncertainty associated with isotopic prediction in spent fuel assemblies for burnup credit analysis. The analysis applies to the design of criticality control systems for spent fuel casks. A total of 54 spent fuel samples were modeled and analyzed using the Shielding Analyses Sequence (SAS2H). Multiple regression analysis and a trending test were performed to develop isotopic correction factors for 10 actinide burnup credit isotopes. 5 refs., 1 tab.

  16. Isotopic and criticality validation for actinide-only burnup credit

    SciTech Connect

    Fuentes, E.; Lancaster, D.; Rahimi, M.

    1997-07-01

    The techniques used for actinide-only burnup credit isotopic validation and criticality validation are presented and discussed. Trending analyses have been incorporated into both methodologies, requiring biases and uncertainties to be treated as a function of the trending parameters. The isotopic validation is demonstrated using the SAS2H module of SCALE 4.2, with the 27BURNUPLIB cross section library; correction factors are presented for each of the actinides in the burnup credit methodology. For the criticality validation, the demonstration is performed with the CSAS module of SCALE 4.2 and the 27BURNUPLIB, resulting in a validated upper safety limit.

  17. Edge dislocation core structures in FCC metals determined from ab initio calculations combined with the improved Peierls-Nabarro equation

    Microsoft Academic Search

    Rui Wang; Shaofeng Wang; Xiaozhi Wu

    2011-01-01

    We have employed the improved Peierls-Nabarro (P-N) equation to study the properties of 1\\/2lang110rang edge dislocation in the {111} plane in face-centered cubic (FCC) metals Al, Cu, Ir, Pd and Pt. The generalized-stacking-fault energy surface entering the equation is calculated by using first-principles density functional theory (DFT). The accuracy of the method has been tested by calculating the values for

  18. The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

    SciTech Connect

    Mitenkova, E. F.; Novikov, N. V. [Nuclear Safety Inst. of Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Blokhin, A. I. [State Scientific Center of Russian Federation, Inst. of Physics and Power Engineering Named after A.I. Leypunsky, Bondarenko Square 1, Obninsk, Kaluga Region, 249030 (Russian Federation)

    2012-07-01

    The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

  19. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  20. Methodologies to assess potential lifetime limits for extended burnup nuclear fuel 

    E-print Network

    De Vore, Curtis Vincent

    1986-01-01

    AND OBJECTIVES. CHAPTER II. HIGH BURNUP DATA DISTRIBUTION INTRODUCTION METHODOLOGY. RESULTS AND DISCUSSION. 12 23 CONCLUSIONS 71 CHAPTER III. COMBINED HIGH AND LOW BURNUP DATA DISTRIBUTION. INTRODUCTION METHODOLOGY 75 RESULTS AND DISCUSSION. 80... CONCLUSIONS 116 CHAPTER IV. HIGH BURNUP DATA RESPONSE TRENDS. . 121 INTRODUCTION METHODOLOGY. 121 122 RESULTS AND DISCUSSION 123 Page CONCLUSIONS. 152 CHAPTER V. POTENTIAL FOR CLADDING FAILURE DURING TRANSIENT EVENTS AT EXTENDED BURNUPS. . . 159...

  1. Core-valence Gaussian basis sets of double and triple zeta quality for Li to Ar. Applications in calculations of indirect nuclear spin-spin coupling constants

    NASA Astrophysics Data System (ADS)

    de Oliveira, P. J. P.; Gomes, M. S.; Pires, J. M.

    2012-09-01

    In this Letter we extend the XZP basis sets (X = D and T) developed by Jorge et al. for Li-Ar atoms with the tight functions and optimize these functions using the criterion of maximizing the core correlation energy (CCE) developed by Woon and Dunning. The basis sets generated with this method were designated as CXZP. Our results showed CCE of the CXZP sets compared to the XZP sets were between 153 and 240 millihartrees (for Na-Ar). Applications in calculations of NMR indirect spin-spin coupling constants at the B3LYP and SOPPA levels were performed.

  2. Impacts of the use of spent nuclear fuel burnup credit on DOE advanced technology legal weight truck cask GA-4 fleet size

    SciTech Connect

    Mobasheran, A.S. [Roy F. Weston, Inc., Washington, DC (United States); Boshoven, J. [General Atomics, San Diego, CA (United States); Lake, B. [Department of Energy, Washington, DC (United States)

    1995-12-01

    The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit.

  3. 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. [Dept. of Nuclear and Theoretical Physics, Univ. of Pavia, 27100 Pavia (Italy); Italian National Inst. of Nuclear Physics INFN, Section of Pavia, Via A. Bassi, 6, 27100 Pavia (Italy); Alloni, D.; Cagnazzo, M.; Magrotti, G.; Manera, S.; Prata, M.; Salvini, A. [Italian National Inst. of Nuclear Physics INFN, Section of Pavia, Via A. Bassi, 6, 27100 Pavia (Italy); Applied Nuclear Energy Laboratory LENA, Univ. of Pavia, Via Aselli, 41, 27100 Pavia (Italy); Chiesa, D.; Clemenza, M.; Pattavina, L.; Previtali, E.; Sisti, M. [Physics Dept. G. Occhialini, Univ. of Milano Bicocca, 20126 Milano (Italy); Italian National Inst. of Nuclear Physics INFN, Section of Milano Bicocca, P.zza della Scienza, 3, 20126 Milano (Italy); Cammi, A. [Italian National Inst. of Nuclear Physics INFN, Section of Milano Bicocca, P.zza della Scienza, 3, 20126 Milano (Italy); Dept. of Energy Enrico Fermi Centre for Nuclear Studies CeSNEF, Polytechnic Univ. of Milan, Via U. Bassi, 34/3, 20100 Milano (Italy)

    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)

  4. First-principles core-level X-ray photoelectron spectroscopy calculation on arsenic defects in silicon crystal

    SciTech Connect

    Kishi, Hiroki; Miyazawa, Miki; Matsushima, Naoki; Yamauchi, Jun [Faculty of Science and Technology, Keio University, 3-14-1 Hiyoshi, Yokohama-shi, Kanagawa-ken 223-8522 (Japan)

    2014-02-21

    We investigate the X-ray photoelectron spectroscopy (XPS) binding energies of As 3d in Si for various defects in neutral and charged states by first-principles calculation. It is found that the complexes of a substitutional As and a vacancy in charged and neutral states explain the experimentally observed unknown peak very well.

  5. REVIEWS OF TOPICAL PROBLEMS: Viscosity measurements on metal melts at high pressure and viscosity calculations for the earth's core

    Microsoft Academic Search

    Vladimir N. Mineev; Aleksandr I. Funtikov

    2004-01-01

    A review is given of experimental and calculated data on the viscosity of iron-based melts on the melting curve. The interest in these data originates in the division of opinion on whether viscosity increases rather moderately or considerably in the high-pressure range. This disagreement is especially pronounced in the interpretation of the values of molten iron and its compounds in

  6. 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.

  7. Study on Ultra-Long Life,Small U-Zr Metallic Fuelled Core With Burnable Poison

    SciTech Connect

    Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa [Chubu Electric Power Company Inc., 1, Higashi-shincho Higashi-ku, Nagoya-shi, ACH 461-8680 (Japan); Hisato Matsumiya; Yoshiaki Sakashita; Yasuyuki Moriki; Mitsuaki Yamaoka; Norihiko Handa [Toshiba Corporation (Japan)

    2002-07-01

    A conceptual design for a 50 MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled core has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up. (authors)

  8. Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

    PubMed

    Dawahra, S; Khattab, K; Saba, G

    2015-07-01

    Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39mk) and after the reactor shutdown (0.735mk), It followed by U3Si-Al (0.34mk, 0.653mk), U3Si2-Al (0.33mk, 0.634mk), U9Mo-Al (0.3mk, 0.568mk) and UO2 (0.24mk, 0.448mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. PMID:25816783

  9. 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 burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

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

    SciTech Connect

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

    2012-07-01

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

  11. Mathematical modeling of the heat treatment and combustion of a coal particle. V. Burn-up stage

    NASA Astrophysics Data System (ADS)

    Enkhjargal, Kh.; Salomatov, V. V.

    2011-07-01

    The present material is a sequel of the previous publications of the authors in this journal under a common title in which by means of mathematical modeling the sequential stages of the process of combustion of coal fuels have been obtained: heating, drying, escape of volatiles, and ignition. Mathematical models of the final stage of combustion of an individual particle — the burn-up stage — have been formulated. On the basis of the solution methods for nonlinear boundary-value problems developed by us, approximate-analytic formulas for two characteristic regimes, burn-up simultaneously with the evaporation of the remaining moisture and burn-up of the completely dried coke residue, have been obtained. The previous history of the physical and chemical phenomena in the general burning pattern is taken into account. The influence of the ash shell on the duration of combustion has been extimated. Comparison of calculations by the obtained dependences with the results of other authors has been made. It showed an accuracy sufficient for engineering applications.

  12. Orbital-resolved vortex-core states in FeSe superconductors: A calculation based on a three-orbital model

    NASA Astrophysics Data System (ADS)

    Wang, Q. E.; Zhang, F. C.

    2015-06-01

    We study the electronic structure of vortex core states of FeSe superconductors based on a t2 g three-orbital model by solving the Bogoliubov-de Gennes (BdG) equation self-consistently. The orbital-resolved vortex core states of different pairing symmetries manifest themselves as distinguishable structures due to different quasiparticle wave functions. The obtained vortices are classified in terms of the invariant subgroups of the symmetry group of the mean-field Hamiltonian in the presence of a magnetic field. Isotropic s and anisotropic s -wave vortices have G5 symmetry for each orbital, whereas dx2-y2-wave vortices show G6* symmetry for dx z /y z orbitals and G5* symmetry for dx y orbital. In the case of dx2-y2-wave vortices, hybridized-pairing between dx z and dy z orbitals gives rise to a relative phase difference in terms of gauge transformed pairing order parameters between dx z /y z and dx y orbitals, which is essentially caused by a transformation of co-representation of G5* and G6* subgroups. The calculated local density of states (LDOS) of dx2-y2-wave vortices shows a qualitatively similar pattern with the experimental results. The phase difference of ?/4 between dx z /y z and dx y orbital-resolved dx2-y2-wave vortices can be verified by further experimental observation.

  13. 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.

  14. Global variance reduction for Monte Carlo reactor physics calculations

    SciTech Connect

    Zhang, Q.; Abdel-Khalik, H. S. [Department of Nuclear Engineering, North Carolina State University, P.O. Box 7909, Raleigh, NC 27695-7909 (United States)

    2013-07-01

    Over the past few decades, hybrid Monte-Carlo-Deterministic (MC-DT) techniques have been mostly focusing on the development of techniques primarily with shielding applications in mind, i.e. problems featuring a limited number of responses. This paper focuses on the application of a new hybrid MC-DT technique: the SUBSPACE method, for reactor analysis calculation. The SUBSPACE method is designed to overcome the lack of efficiency that hampers the application of MC methods in routine analysis calculations on the assembly level where typically one needs to execute the flux solver in the order of 10{sup 3}-10{sup 5} times. It places high premium on attaining high computational efficiency for reactor analysis application by identifying and capitalizing on the existing correlations between responses of interest. This paper places particular emphasis on using the SUBSPACE method for preparing homogenized few-group cross section sets on the assembly level for subsequent use in full-core diffusion calculations. A BWR assembly model is employed to calculate homogenized few-group cross sections for different burn-up steps. It is found that using the SUBSPACE method significant speedup can be achieved over the state of the art FW-CADIS method. While the presented speed-up alone is not sufficient to render the MC method competitive with the DT method, we believe this work will become a major step on the way of leveraging the accuracy of MC calculations for assembly calculations. (authors)

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

    SciTech Connect

    Su'ud, Zaki, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Sekimoto, H., E-mail: hsekimot@gmail.com [Research Lab. For Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo (Japan)

    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.

  16. Gamma-Ray Simulated Spectrum Deconvolution of a LaBr? 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 LaBr? 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” LaBr? 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

  17. Natural Nuclear Reactor Oklo and Variation of Fundamental Constants Part 1: Computation of Neutronics of Fresh Core

    E-print Network

    Yu. V. Petrov; A. I. Nazarov; M. S. Onegin; V. Yu. Petrov; E. G. Sakhnovsky

    2005-09-15

    Using modern methods of reactor physics we have performed full-scale calculations of the natural reactor Oklo. For reliability we have used recent version of two Monte Carlo codes: Russian code MCU REA and world wide known code MCNP (USA). Both codes produce similar results. We have constructed a computer model of the reactor Oklo zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities and neutron fluxes were calculated. We have estimated also the temperature and void effects for the fresh core. As would be expected, we have found for the fresh core a significant difference between reactor and Maxwell spectra, which was used before for averaging cross sections in the Oklo reactor. The averaged cross section of Sm-149 and its dependence on the shift of resonance position (due to variation of fundamental constants) are significantly different from previous results. Contrary to results of some previous papers we find no evidence for the change of the fine structure constant in the past and obtain new, most accurate limits on its variation with time: -4 10^{-17}year^{-1} < d alpha/dt/alpha < 3 10^{-17} year^{-1} A further improvement in the accuracy of the limits can be achieved by taking account of the core burnup. These calculations are in progress.

  18. 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 Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D2O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D2O-HCPWR.

  19. On the oxidation state of UO 2 nuclear fuel at a burn-up of around 100 MWd\\/kgHM

    Microsoft Academic Search

    C. T. Walker; V. V. Rondinella; D. Papaioannou; S. Van Winckel; W. Goll; R. Manzel

    2005-01-01

    Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102MWd\\/kgHM are presented. The local ?G¯(O2) of the fuel was measured using a miniature solid state galvanic cell, the local O\\/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O\\/M ratio was derived

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

    SciTech Connect

    Enercon Services, Inc.

    2011-03-14

    Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in 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 compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the cos

  1. Attainable Burnup in a LIFE Engine Loaded with Depleted Uranium

    SciTech Connect

    Fratoni, M; Kramer, K J; Latkowski, J F

    2009-11-30

    The Laser Inertial Fusion-based Energy (LIFE) system uses a laser-based fusion source for electricity production. The (D,T) reaction, beside a pure fusion system, allows the option to drive a sub-critical fission blanket in order to increase the total energy gain. In a typical fusion-fission LIFE engine the fission blanket is a spherical shell around the fusion source, preceded by a beryllium shell for neutron multiplications by means of (n,2n) reactions. The fuel is in the form of TRISO particles dispersed in carbon pebbles, cooled by flibe. The optimal design features 80 cm thick blanket, 16 cm multiplier, and 20% TRISO packing factor. A blanket loaded with depleted uranium and depleted in a single batch with continuous mixing can achieve burnup as high as {approx}85% FIMA while generating 2,000 MW of total thermal power and producing enough tritium to be used for fusion. A multi-segment blanket with a central promotion shuffling scheme enhances burnup to {approx}90% FIMA, whereas a blanket that is operated with continuous refueling achieves only 82% FIMA under the same constraints of thermal power and tritium self-sufficiency. Both, multi-segment and continuous refueling eliminate the need for a fissile breeding phase.

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

    Microsoft Academic Search

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

    2003-01-01

    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

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

    Microsoft Academic Search

    F. Lemoine

    1997-01-01

    Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning

  4. Next Generation CANDU Core Physics Innovations

    SciTech Connect

    Chan, P.S.W.; Hopwood, J.M.; Love, J.W. [Atomic Energy of Canada Ltd., Ontario (Canada)

    2002-07-01

    NG CANDU is the 'Next Generation' CANDU{sup R} reactor, aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs. A key element of cost reduction is the use of H{sub 2}O as coolant and Slightly Enriched Uranium fuel in a tight D{sub 2}O-moderated lattice. The innovations in the CANDU core physics result in substantial improvements in economics as well as significant enhancements in reactor licensability, controllability, and waste reduction. The full-core coolant-void reactivity in NG CANDU is about -3 mk. Power coefficient is substantially negative. Fuel burnup is about three times the current natural-uranium burnup. (authors)

  5. Features of the application of the Monte Carlo method to calculations for large RBMK reactors and to model correction on the basis of data from in-core detectors

    SciTech Connect

    Ivanov, I. E., E-mail: ilshai-hulud@yandex.ru; Schukin, N. V. [National Research Nuclear University MEPhI (Russian Federation); Bychkov, S. A.; Druzhinin, V. E.; Lysov, D. A.; Shmonin, Yu. V. [All-Russia Research Institute for Nuclear Power Plant Operation (VNIIAES) (Russian Federation); Gurevich, M. I. [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15

    Statistical errors in sampling neutron fields in physically large systems like an RBMK are analyzed both qualitatively and quantitatively. Recommendations concerning the choice of parameters for calculations are given. A new procedure for Monte Carlo RBMK calculations with model corrections on the basis of data from in-core detectors is proposed. Dedicated software based on the CUDA software and hardware platform is developed for computational research. Results of testing the procedure and software in question via calculations for real RBMK reactors are discussed.

  6. High burnup performance of an advanced oxide fuel assembly in FFTF (Fast Flux Test Facility) with ferritic/martensitic materials

    SciTech Connect

    Bridges, A.E.; Saito, G.H.; Lovell, A.J.; Makenas, B.J.

    1986-05-01

    An advanced oxide fuel assembly with ferritic/martensitic materials has successfully completed its sixth cycle of irradiation in the FFTF, reaching a peak pellet burnup greater than 100 MWd/KgM and a peak fast fluence greater than 15 x 10SS n/cmS. The cladding, wire-wrap, and duct material for the ACO-1 test assembly is the ferritic/martensitic alloy, HT9, which was chosen for use in long-lifetime fuel assemblies because of its good nominal temperature creep strength and low swelling rate. Valuable experience on the performance of HT9 materials has been gained from this test, advancing our quest for long-lifetime fuel. Pertinent data, obtained from the ACO-1 test assembly, will support the irradiation of the Core Demonstration Experiment in FFTF.

  7. Analytical core loss calculations for magnetic materials used in high frequency high power converter applications. Ph.D. Thesis - Toledo Univ.

    NASA Technical Reports Server (NTRS)

    Triner, J. E.

    1979-01-01

    The basic magnetic properties under various operating conditions encountered in the state-of-the-art DC-AC/DC converters are examined. Using a novel core excitation circuit, the basic B-H and loss characteristics of various core materials may be observed as a function of circuit configuration, frequency of operation, input voltage, and pulse-width modulation conditions. From this empirical data, a mathematical loss characteristics equation is developed to analytically predict the specific core loss of several magnetic materials under various waveform excitation conditions.

  8. Assessment of the use of extended burnup fuel in light water power reactors

    SciTech Connect

    Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

    1988-02-01

    This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

  9. A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

    SciTech Connect

    Yang, W.S.; Kim, T.K.; Grandy, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne IL 60439 (United States)

    2007-07-01

    This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% {delta}k. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% {delta}k. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

  10. Modified Laser and Thermos cell calculations on microcomputers

    Microsoft Academic Search

    A. Shapiro; H. C. Huria

    1987-01-01

    In the course of designing and operating nuclear reactors, many fuel pin cell calculations are required to obtain homogenized cell cross sections as a function of burnup. In the interest of convenience and cost, it would be very desirable to be able to make such calculations on microcomputers. In addition, such a microcomputer code would be very helpful for educational

  11. Comparison of XSUSA and 'two-step' approaches for full-core uncertainty quantification

    SciTech Connect

    Yankov, A. [Univ. of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Klein, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH, Boltzmannstr. 14, D- 85748 Garching b. Muenchen (Germany); Jessee, M. A. [Oak Ridge National Laboratory (United States); Zwermann, W.; Velkov, K.; Pautz, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH, Boltzmannstr. 14, D- 85748 Garching b. Muenchen (Germany); Collins, B.; Downar, T. [Univ. of Michigan (United States)

    2012-07-01

    While there are multiple sources of error that are introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily for quantifying the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the 'Two-Step' method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the Two-Step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution will be examined in the framework of phase 1-3 of the UAM Benchmark. Using the TMI core as a base model for analysis, the XSUSA and Two-Step methods are applied with certain limitations and the results are compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions are made for which method is currently a more viable option for computing uncertainties in burnup and transient calculations. (authors)

  12. Comparison of XSUSA and "Two-Step" Approaches for Full Core Uncertainty Quantification

    SciTech Connect

    Yankov, Artem [University of Michigan; Klein, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Jessee, Matthew Anderson [ORNL; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Velkov, Kiril [Gesellschaft fur Anlagen; Pautz, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Collins, Benjamin [University of Michigan; Downar, Thomas [University of Michigan

    2012-01-01

    While there are multiple sources of error that are introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily for quantifying the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the 'Two-Step' method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the Two-Step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution will be examined in the framework of phase I-3 of the UAM Benchmark. Using the TMI core as a base model for analysis, the XSUSA and Two-Step methods are applied with certain limitations and the results are compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions are made for which method is currently a more viable option for computing uncertainties in burnup and transient calculations.

  13. Construction and utilization of linear empirical core models for PWR in-core fuel management

    SciTech Connect

    Okafor, K.C.

    1988-01-01

    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k{sub {infinity}} profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results.

  14. calculators hp calculators

    E-print Network

    Vetter, Frederick J.

    calculators hp calculators HP 50g Calculations involving plots Plotting on the HP 50g The 2D/3D;hp calculators HP 50g Calculations involving plots hp calculators - 2 - HP 50g Calculations involving plots Plotting on the HP 50g The HP 50g calculator provides a host of plots to allow the user

  15. Assessment of high-burnup LWR fuel response to reactivity-initiated accidents

    E-print Network

    Liu, Wenfeng, Ph.D. Massachusetts Institute of Technology

    2007-01-01

    The economic advantages of longer fuel cycle, improved fuel utilization and reduced spent fuel storage have been driving the nuclear industry to pursue higher discharge burnup of Light Water Reactor (LWR) fuel. A design ...

  16. Design strategies for optimizing high burnup fuel in pressurized water reactors

    E-print Network

    Xu, Zhiwen, 1975-

    2003-01-01

    This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

  17. Gamma-Ray Simulated Spectrum Deconvolution of a LaBr3 1- ×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 bumup 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

  18. Gamma-Ray Simulated Spectrum Deconvolution of a LaBr3 1- ×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 bumup 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.

  19. A Simplified Approach for Evaluation of the Burnup Potential of Alternative Fuels

    Microsoft Academic Search

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

    2004-01-01

    To evaluate the burnup potential of a fuel pin, a simplified fuel rod analysis code called FUEL-SIMplified model (FUELSIM) was developed using the general-purpose software VENSIM. FUELSIM is based on FRAPCON-3 models and validated against it. A sensitivity analysis was done using FUELSIM to determine the fuel parameters that have high importance in limiting the burnup potential of a fuel

  20. Extended burnup demonstration reactor fuel program. Semi-annual progress report, January 1979-September 1979

    SciTech Connect

    Woods, K.N.; van Swam, L.F.

    1980-12-31

    This is the first semi-annual progress report for the DOE-sponsored Extended Burnup Demonstration program. The program objectives, description, and organization are detailed. Characteristics are given for the 64 Big Rock Point fuel rods and the four 8 x 8 Oyster Creek fuel assemblies which will be driven to extended burnup. The transfer of 64 Big Rock Point fuel rods from their original assemblies into host assemblies and the results of the fuel examination of these rods are described.

  1. Measurement and calculation of high-actinide burnup in the prototype fast reactor

    Microsoft Academic Search

    B. L. Broadhead; S. Raman; J. K. Dickens

    1991-01-01

    An agreement was signed in May 1979 as a part of a long-term cooperative program between the United Kingdom and the US under the liquid-metal fast breeder reactor agreement of 1976. This agreement included an experiment to carry out irradiations of physics specimens of fissile and fertile actinides to improve our knowledge of basic nuclear physics phenomena. Three fuel pins

  2. RIS-M-2185 CALCULATION OF HEAT RATING AND BURN-UP FOR TEST FUEL PINS

    E-print Network

    , FLEI PELLETS, FUEL PINS, MATHEMATICAL MODELS, POWER DISTRIBUTION, RADIATION HEATING, URANIUM DIOXIDE element initially contains highly enriched U-235 (120 or 150 g) alloyed with and clad in aluminium

  3. Potential of pin-by-pin SPN calculations as an industrial reference

    SciTech Connect

    Fliscounakis, M.; Girardi, E.; Courau, T.; Couyras, D. [EDF R and D/Sinetics, 1 av du General de Gaulle, F92141 Clamart Cedex (France)

    2012-07-01

    This paper aims at analysing the potential of pin-by-pin SP{sub n} calculations to compute the neutronic flux in PWR cores as an alternative to the diffusion approximation. As far as pin-by-pin calculations are concerned, a SPH equivalence is used to preserve the reactions rates. The use of SPH equivalence is a common practice in core diffusion calculations. In this paper, a methodology to generalize the equivalence procedure in the SP{sub n} equations context is presented. In order to verify and validate the equivalence procedure, SP{sub n} calculations are compared to 2D transport reference results obtained with the APOLL02 code. The validation cases consist in 3x3 analytical assembly color sets involving burn-up heterogeneities, UOX/MOX interfaces, and control rods. Considering various energy discretizations (up to 26 groups) and flux development orders (up to 7) for the SP{sub n} equations, results show that 26-group SP{sub 3} calculations are very close to the transport reference (with pin production rates discrepancies < 1%). This proves the high interest of pin-by-pin SP{sub n} calculations as an industrial reference when relying on 26 energy groups combined with SP{sub 3} flux development order. Additionally, the SP{sub n} results are compared to diffusion pin-by-pin calculations, in order to evaluate the potential benefit of using a SP{sub n} solver as an alternative to diffusion. Discrepancies on pin-production rates are less than 1.6% for 6-group SP{sub 3} calculations against 3.2% for 2-group diffusion calculations. This shows that SP{sub n} solvers may be considered as an alternative to multigroup diffusion. (authors)

  4. Sensitivity study on Xe depletion in the high burn-up structure of UO2

    NASA Astrophysics Data System (ADS)

    Holt, L.; Schubert, A.; Van Uffelen, P.; Walker, C. T.; Fridman, E.; Sonoda, T.

    2014-09-01

    Experimental results for the Xe depletion in the matrix of high burn-up fuel are presented from the High Burnup Rim Project (HBRP). In this project a number of UO2 fuel discs with 235U enrichment of 25.8 wt.% were irradiated. The Xe content of the fuel discs was analysed by means of electron probe microanalysis (EPMA). The influence of the burn-up and irradiation temperature on the Xe concentration was investigated using a multi-physics approach involving various simulation tools. The temperature influence was modelled by means of the temperature dependent effective burn-up. Good agreement was found between the modelled temperature threshold of the effective burn-up and the experimental temperature threshold between un- and restructured fuel in the HBRP. However, a systematic difference is observed between the onset burn-up derived from the Xe measurements in highly enriched discs such as those of HBRP and the corresponding values derived from irradiated Light Water Reactor (LWR) fuel rods and reported in the open literature. A sensitivity study identified the neutron flux spectrum and the fission product yields as the main reasons for the observed differences.

  5. Dissolution of low burnup Fast Flux Test reactor fuel

    SciTech Connect

    Fellows, R.L.; Campbell, D.O.; Mailen, J.C.

    1984-01-01

    The first Fast-Flux Test Facility reactor fuel (mixed (U,Pu)O/sub 2/ composition) has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 1997/sup 0/C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 95/sup 0/C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 29/sup 0/C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 95/sup 0/C dissolution contained the equivalent of 198 mg of /sup 239/Pu per 100 g of hulls, while the cladding from the 29/sup 0/c experiments contained only 0.21 mg of /sup 239/Pu per 100 g of hulls. 9 references, 5 figures.

  6. Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

    SciTech Connect

    Tiberi, V. [Institut de Radioprotection et de Surete Nucleaire IRSN, PSN-EXP/SNC/LNR, BP 17, 92262 Fontenay-aux-Roses (France)

    2012-07-01

    The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity of the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)

  7. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    Microsoft Academic Search

    Olena Gritzay; Oleksandr Kalchenko; Nataliya Klimova; Volodymyr Razbudey; Andriy Sanzhur; Stephen Binney

    2005-01-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor

  8. Core Optimization of a Deep-Burn Pebble Bed Reactor

    SciTech Connect

    Brian Boer; Abderrafi M. Ougouag

    2010-06-01

    Achieving a high fuel burnup in the Deep-Burn (DB) pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum as compared to a ’standard’ UO2 fueled core. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. The DB concept focuses on the destruction of spent fuel transuranics in TRISO coated particle fueled gas-cooled reactors with the aim of a fractional fuel burnup of 60-70% in fissions per initial metal atom (FIMA), using a single-pass, multi in-core fuel (re)cycling scheme. In principle, the DB pebble bed concept employs the same reactor designs as the present low enriched uranium core designs, i.e. the 400 MWth Pebble Bed Modular Reactor (PBMR-400). A Pu and Minor Actinide fueled PBMR-400 design serves as the starting point for a core optimization study. The fuel temperature, power peak, temperature reactivity coefficients, and burnup capabilities of the modified designs are analyzed with the PEBBED code. A code-to-code coupling with the PASTA code allows for the analysis of the TRISO fuel performance for both normal and Loss Of Forced Cooling conditions. An improved core design is sought, maximizing the fuel discharge burnup, while retaining negative temperature reactivity feedback coefficients for the entire temperature range and avoiding high fuel temperatures (fuel failure probabilities).

  9. Construction of linear empirical core models for pressurized water reactor in-core fuel management

    SciTech Connect

    Okafor, K.C.; Aldemir, T. (The Ohio State Univ., Dept. of Mechanical Engineering, Nuclear Engineering Program, 206 West 18th Ave., Columbus, OH (US))

    1988-06-01

    An empirical core model construction procedure for pressurized water reactor (PWR) in-core fuel management problems is presented that (a) incorporates the effect of composition changes in all the control zones in the core of a given fuel assembly, (b) is valid at all times during the cycle for a given range of control variables, (c) allows determining the optimal beginning of cycle (BOC) kappainfinity distribution as a single linear programming problem,and (d) provides flexibility in the choice of the material zones to describe core composition. Although the modeling procedure assumes zero BOC burnup, the predicted optimal kappainfinity profiles are also applicable to reload cores. In model construction, assembly power fractions and burnup increments during the cycle are regarded as the state (i.e., dependent) variables. Zone enrichments are the control (i.e., independent) variables. The model construction procedure is validated and implemented for the initial core of a PWR to determine the optimal BOC kappainfinity profiles for two three-zone scatter loading schemes. The predicted BOC kappainfinity profiles agree with the results of other investigators obtained by different modeling techniques.

  10. Shift of x-ray-photoelectron core levels in Bi sub 2 Sr sub 2 Ca sub 1 minus x Y sub x Cu sub 2 O sub y : An explanation by bond-valence-sum calculation

    SciTech Connect

    Itti, R.; Munakata, F.; Ikeda, K.; Yamauchi, H.; Koshizuka, N.; Tanaka, S. (Superconductivity Research Laboratory, International Superconductivity Technology Center, 1-10-13 Shinonome, Koto-ku, Tokyo 135, Japan (JP))

    1991-03-01

    Shifts of the binding energies of x-ray-photoelectron core levels are observed in the Bi{sub 2}Sr{sub 2}Ca{sub 1{minus}{ital x}}Y{sub {ital x}}Cu{sub 2}O{sub {ital y}} system when {ital x} is varied. It is pointed out that this is due to the change in the chemical environment associated with the changes of the interatomic distances. We show that the so-called bond-valence-sum calculation is useful for the explanation of the shift; and propose that this method is generally applicable in explaining or predicting the shift of core levels in any system with a variable composition.

  11. calculators hp calculators

    E-print Network

    Vetter, Frederick J.

    calculators hp calculators HP 50g Working with Parametric Plots Plotting on the HP 50g Parametric calculators HP 50g Working with Parametric Plots hp calculators - 2 - HP 50g Working with Parametric Plots Plotting on the HP 50g The HP 50g calculator provides a host of plots to allow the user to visualize data

  12. A Simplified Approach for Evaluation of the Burnup Potential of Alternative Fuels

    SciTech Connect

    Oggianu, Stella Maris [Massachusetts Institute of Technology (United States); No, Hee Cheon [Korea Advanced Institute of Science and Technology (Korea, Republic of); Kazimi, Mujid S. [Massachusetts Institute of Technology (United States)

    2004-06-15

    To evaluate the burnup potential of a fuel pin, a simplified fuel rod analysis code called FUEL-SIMplified model (FUELSIM) was developed using the general-purpose software VENSIM. FUELSIM is based on FRAPCON-3 models and validated against it. A sensitivity analysis was done using FUELSIM to determine the fuel parameters that have high importance in limiting the burnup potential of a fuel material. Among 16 parameters, 10 were identified as having high importance. For six fuel materials (uranium metal, UC, UN, Th/U metal, UO{sub 2}/ThO{sub 2} fuels, and UO{sub 2}), a simplified model for the pressure rise and volumetric changes inside the fuel is developed to estimate the operational index of each fuel; these models include only the variables with high importance. It was found that the highest burnup potential is that of the nitride fuel, followed by the UO{sub 2}/ThO{sub 2} fuel.

  13. Preparation and characterization of the simulated burnup americium-containing uranium-plutonium mixed oxide fuel

    NASA Astrophysics Data System (ADS)

    Tanaka, Kosuke; Osaka, Masahiko; Miwa, Shuhei; Hirosawa, Takashi; Kurosaki, Ken; Muta, Hiroaki; Uno, Masayoshi; Yamanaka, Shinsuke

    2012-01-01

    In order to investigate the effect on fuel thermophysical properties when adding americium and selected fission products to uranium-plutonium mixed oxide (MOX) fuel, simulated low decontamination MOX fuel with high burn-ups to 250 GWd/t, has been prepared and subjected to characterization tests, elastic moduli measurements and melting temperature measurement. Elastic moduli for the simulated low decontamination MOX fuel were almost the same level as fuel without americium and fission products and decrease in the moduli was slight with increasing simulated burn-up. The melting temperature of high burn-up, low decontamination MOX fuel may be estimated by using the findings on the effect of americium, plutonium addition and fission products accumulation.

  14. Role of Minor Actinides for Long-Life Reactor Cores

    SciTech Connect

    Saito, M.; Artisyuk, V. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152 (Japan); Shmelev, A. [Moscow Engineering and Physics Institute, 31, Kashirskoe Shosse, Moscow (Russian Federation); Nikitin, K.; Peryoga, Y

    2002-07-01

    The paper addresses the study on advanced fuel cycles for LWR oriented to high burnup values that exceed 100 GWd/tHM, thus giving the chance to establish the long-life reactor cores without fuel reloading on site. The key element of this approach is a broad involvement of Minor Actinides whose admixture to 20% enriched uranium fuel provides safe release of initial reactivity excess and improved proliferation resistance properties. (authors)

  15. calculators hp calculators

    E-print Network

    Vetter, Frederick J.

    calculators hp calculators HP 50g The basics of plotting functions Plotting on the HP 50g The 2D/3D (PLOT SETUP) Form The WIN Form Examples of plotting functions #12;hp calculators HP 50g The basics of plotting functions hp calculators - 2 - HP 50g The basics of plotting functions Plotting on the HP 50g

  16. Development of a new measurement method for fast breeder reactor fuel burnup using a shielded ion microprobe analyzer

    Microsoft Academic Search

    M. Mizuno; Y. Enokido; T. Itaki; K. Kono; I. Unno; S. Yamanouchi

    1985-01-01

    A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of

  17. A search for minimum volume of Breed and Burn cores

    SciTech Connect

    Di Sanzo, C.; Greenspan, E. [Dept. of Nuclear Engineering, Univ. of California, Berkeley Etcheverry Hall, Berkeley, CA 94720 (United States)

    2012-07-01

    The objective of the present study is to quantify the minimum volume a Breed and Burn (B and B) core can be designed to have and the corresponding burnup required for sustaining the breed-and-burn mode of operation based on neutronics; radiation damage constraints are ignored. The minimum radius for an idealized spherical B and B reactor is 136 cm or 110 cm for, respectively, 40% or 28% coolant volume fraction. The peak required burnup is about 25%. The minimum volume of a more realistic cylindrical B and B core is estimated to be only {approx}15% larger than that of the idealized spherical core but is only 43% of the volume of the medium-size B and B core previously designed to fit within the S-Prism reactor vessel. Thus it appears that SMR s can, in principle, be designed to have a B and B core. It was also found that the minimum volume B and B core does not necessarily coincide with the maximum permissible leakage from a core that can sustain the B and B mode of operation. (authors)

  18. TRU transmutation in thorium-based heterogeneous PWR core

    SciTech Connect

    Bae, Kang-Mok; Lim, Jae-Yong; Kim, Myung-Hyun [Department of Nuclear Engineering, Kyung Hee University, YoungIn-shi, Gyeonggi-do, 449-701 (Korea, Republic of)

    2004-07-01

    A thorium-based seed and blanket design concept for a conventional pressurized light water reactor (PWR) was proposed to enhance the proliferation resistance potential and fuel cycle economics. The KTF core was satisfied with neutronic and thermal-hydraulic design limit of conventional PWR, APR-1400. In order to evaluate transmutation capability of a thorium-based KTF core, U/Zr seed fuel mixed with 10% TRU which come from 1,000 MWe power reactor after 10 years decay was proposed and analyzed by transmutation indices such as D{sub j}, TEX and SR. KTF core showed an extended fuel cycle burnup; average burnup of seed was 79.5 MWd/kgHM and blanket was 94.6 MWd/kgHM. It means that residence time of TRU in the core could be long enough for transmutation when TRU is mixed in seed fuel. The amount of TRU production from conventional PWR could be transmuted in the KTF-TRU core, especially Am-241 isotope is remarkably transmuted by capture reaction. Even isotopes of curium were cumulated in the core during the burnup, however, KTF-TRU core could reduce the TRU in spent fuel by using well-thermalized neutron spectrum. Proliferation resistance potential of thorium based transmutation fuel is slightly increased. About 31% reduction of TRU amount was measured from reduced plutonium production from U-238. Total amount of Am-241 was reduced significantly, but total amount of minor actinide (MA) was reduced by 28% of its initial loading mass. (authors)

  19. Flux reconstruction methods for assembly calculations in the code APOLLO2

    SciTech Connect

    Zmijarevic, I.; Masiello, E.; Sanchez, R. [Commissariat a l'Energie Atomique CEA, Direction de l'Energie Nucleaire DEN, DM2S/SERMA, 91191 Gif-sur-Yvette cedex (France)

    2006-07-01

    A technique for flux reconstruction has been incorporated in the code APOLLO2 allowing for fast generation of accurate burnup libraries. The burnup flux is obtained as the product of a pivot flux from a large-macro-group heterogeneous-assembly calculation times an energy shape factor based on a fine-energy flux. The latter is obtained from a fast multicell calculation for a few types of representative cells in the assembly. The reconstruction formula preserves the reaction rates predicted by the large-macro-group, heterogeneous-assembly calculation. Analysis of a BWR MOX benchmark shows that the reactivity error for a 70 MWd/t burnup cycle did not exceed 90 pcm and the maximum error in pin powers did not exceed 1%. (authors)

  20. Modeling the performance of high burnup thoria and urania PWR fuel

    E-print Network

    Long, Yun, 1972-

    2002-01-01

    Fuel performance models have been developed to assess the performance of ThO?-UO? fuels that can be operated to a high burnup up to 80-100MWd/kgHM in current and future Light Water Reactors (LWRs). Among the various issues ...

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

    Microsoft Academic Search

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

    2007-01-01

    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,

  2. Toward very high burnups, a strategy for plutonium utilization in pressurized water reactors

    Microsoft Academic Search

    J. Porta; J.-Y Doriath

    1999-01-01

    The aim of using plutonium more efficiently in pressurized water reactors has led to objectives of high and very high burnups. The reasons are not only economic, but also related to the optimization of the utilization of fissile material and to increased proliferation resistance. Here are presented the reflections that contributed to the definition of a R&D programme conducted by

  3. A Method for the Burnup Analysis of Power Reactors in Equilibrium Operation Cycles

    Microsoft Academic Search

    Sh?ichir? NAKAMURA

    1966-01-01

    This paper describes a method based on the consistent thermal neutron flux concept for burnup analysis of power reactors in equilibrium operation cycle. Radial flux distributions are discussed, as well as the variation of isotopic densities in the fuel, and considerations are given on control rod distribution for maintaining a consistent thermal flux distribution. The present method can be applied

  4. 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.

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

    SciTech Connect

    Chung, H.M. (Argonne National Lab., IL (USA))

    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.

  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. Development and preliminary verification of the 3D core neutronic code: COCO

    SciTech Connect

    Lu, H.; Mo, K.; Li, W.; Bai, N.; Li, J. [Reactor Design and Fuel Management Research Center, China Nuclear Power Technology Research Inst., 47F/A Jiangsu Bldg., Yitian Road, Futian District, Shenzhen (China)

    2012-07-01

    As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code, the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)

  8. Identifying and bounding uncertainties in nuclear reactor thermal power calculations

    SciTech Connect

    Phillips, J.; Hauser, E.; Estrada, H. [Cameron, 1000 McClaren Woods Drive, Coraopolis, PA 15108 (United States)

    2012-07-01

    Determination of the thermal power generated in the reactor core of a nuclear power plant is a critical element in the safe and economic operation of the plant. Direct measurement of the reactor core thermal power is made using neutron flux instrumentation; however, this instrumentation requires frequent calibration due to changes in the measured flux caused by fuel burn-up, flux pattern changes, and instrumentation drift. To calibrate the nuclear instruments, steam plant calorimetry, a process of performing a heat balance around the nuclear steam supply system, is used. There are four basic elements involved in the calculation of thermal power based on steam plant calorimetry: The mass flow of the feedwater from the power conversion system, the specific enthalpy of that feedwater, the specific enthalpy of the steam delivered to the power conversion system, and other cycle gains and losses. Of these elements, the accuracy of the feedwater mass flow and the feedwater enthalpy, as determined from its temperature and pressure, are typically the largest contributors to the calorimetric calculation uncertainty. Historically, plants have been required to include a margin of 2% in the calculation of the reactor thermal power for the licensed maximum plant output to account for instrumentation uncertainty. The margin is intended to ensure a cushion between operating power and the power for which safety analyses are performed. Use of approved chordal ultrasonic transit-time technology to make the feedwater flow and temperature measurements (in place of traditional differential-pressure- based instruments and resistance temperature detectors [RTDs]) allows for nuclear plant thermal power calculations accurate to 0.3%-0.4% of plant rated power. This improvement in measurement accuracy has allowed many plant operators in the U.S. and around the world to increase plant power output through Measurement Uncertainty Recapture (MUR) up-rates of up to 1.7% of rated power, while also decreasing the probability of significant over-power events. This paper will examine the basic elements involved in calculation of thermal power using ultrasonic transit-time technology and will discuss the criteria for bounding uncertainties associated with each element in order to achieve reactor thermal power calculations to within 0.3% to 0.4%. (authors)

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

    SciTech Connect

    none,

    2014-02-27

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

  10. Adaptive core simulation

    NASA Astrophysics Data System (ADS)

    Abdel-Khalik, Hany Samy

    The work presented in this thesis is a continuation of a master's thesis research project conducted by the author to gain insight into the applicability of inverse methods to developing adaptive simulation capabilities for core physics problems. Use of adaptive simulation is intended to improve the fidelity and robustness of important core attributes predictions such as core power distribution, thermal margins and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e. in-core instrumentations readings, to adapt the simulation in a meaningful way. A meaningful adaption will result in high fidelity and robust adapted core simulators models. To perform adaption, we propose an inverse theory approach in which the multitudes of input data to core simulators, i.e. reactor physics and thermal-hydraulic data, are to be adjusted to improve agreement with measured observables while keeping core simulators models unadapted. At a first glance, devising such adaption for typical core simulators models would render the approach impractical. This follows, since core simulators are based on very demanding computational models, i.e. based on complex physics models with millions of input data and output observables. This would spawn not only several prohibitive challenges but also numerous disparaging concerns. The challenges include the computational burdens of the sensitivity-type calculations required to construct Jacobian operators for the core simulators models. Also, the computational burdens of the uncertainty-type calculations required to estimate the uncertainty information of core simulators input data presents a demanding challenge. The concerns however are mainly related to the reliability of the adjusted input data. We demonstrate that the power of our proposed approach is mainly driven by taking advantage of this unfavorable situation. Our contribution begins with the realization that to obtain numerical solutions to demanding computational models, matrix methods are often employed to produce approximately equivalent discretized computational models that may be manipulated further by computers. The discretized models are described by matrix operators that are often rank-deficient, i.e. ill-posed. We introduce a novel set of matrix algorithms, denoted by Efficient Subspace Methods (ESM), intended to approximate the action of very large, dense, and numerically rank-deficient matrix operators. We demonstrate that significant reductions in both computational and storage burdens can be attained for a typical BWR core simulator adaption problem without compromising the quality of the adaption. We demonstrate robust and high fidelity adaption utilizing a virtual core, e.g. core simulator predicted observables with the virtual core either based upon a modified version of the core simulator whose input data are to be adjusted or an entirely different core simulator. Further, one specific application of ESM is demonstrated, that is being the determination of the uncertainties of important core attributes such as core reactivity and core power distribution due to the available ENDF/B cross-sections uncertainties. The use of ESM is however not limited to adaptive core simulation techniques only, but a wide range of engineering applications may easily benefit from the introduced algorithms, e.g. machine learning and information retrieval techniques highly depends on finding low rank approximations to large scale matrices. In the appendix, we present a stand-alone paper that presents a generalized framework for ESM, including the mathematical theory behind the algorithms and several demonstrative applications that are central to many engineering arenas---(a) sensitivity analysis, (b) parameter estimation, and (c) uncertainty analysis. We choose to do so to allow other engineers, applied mathematicians, and scientists from other scientific disciplines to take direct advantage of ESM without having to sail across the sea of reactor core calculations.

  11. Pre-conceptual design study of ASTRID core

    SciTech Connect

    Varaine, F.; Marsault, P.; Chenaud, M. S.; Bernardin, B.; Conti, A.; Sciora, P.; Venard, C.; Fontaine, B.; Devictor, N.; Martin, L. [Alternative Energies and Atomic Energy Commission, CEA, DEN DER, 13108 St Paul lez Durance (France); Scholer, A. C.; Verrier, D. [AREVA-NP, 10 rue J. Recamier, 69456 Lyon Cedex 06 (France)

    2012-07-01

    In the framework of the ASTRID project at CEA, core design studies are performed at CEA with the AREVA and EDF support. At the stage of the project, pre-conceptual design studies are conducted in accordance with GEN IV reactors criteria, in particularly for safety improvements. An improved safety for a sodium cooled reactor requires revisiting many aspects of the design and is a rather lengthy process in current design approach. Two types of cores are under evaluation, one classical derived from the SFR V2B and one more challenging called CFV (low void effect core) with a large gain on the sodium void effect. The SFR V2b core have the following specifications: a very low burn-up reactivity swing (due to a small cycle reactivity loss) and a reduced sodium void effect with regard to past designs such as the EFR (around 2$ minus). Its performances are an average burn-up of 100 GWd/t, and an internal conversion ratio equal to one given a very good behavior of this core during a control rod withdrawal transient). The CFV with its specific design offers a negative sodium void worth while maintaining core performances. In accordance of ASTRID needs for demonstration those cores are 1500 MWth power (600 MWe). This paper will focus on the CFV pre-conceptual design of the core and S/A, and the performances in terms of safety will be evaluated on different transient scenario like ULOF, in order to assess its intrinsic behavior compared to a more classical design like V2B core. The gap in term of margin to a severe accident due to a loss of flow initiator underlines the potential capability of this type of core to enhance prevention of severe accident in accordance to safety demonstration. (authors)

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

    Microsoft Academic Search

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

    2010-01-01

    The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO2 LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35MWd\\/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded “laser-flash” device and show that the thermal

  13. 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.

  14. 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.

  15. Fission gas release and volume diffusion enthalpy in UO 2 irradiated at low and high burnup

    Microsoft Academic Search

    J. P Hiernaut; C. Ronchi

    2001-01-01

    Samples of UO2, irradiated in LWRs at burnups from 25000 to 95000 MWd\\/t at in-pile temperatures below 800 K, were submitted to Knudsen-effusion experiments. In addition to the equilibrium vapour pressure of non-volatile species, release of fission gas was measured as a function of temperature and time up to approximately 2700 K, where complete vaporisation of the sample was eventually

  16. Fission gas and iodine release measured up to 15 GWd\\/t UOâ burnup

    Microsoft Academic Search

    Appelhans

    1983-01-01

    A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd\\/t UOâ burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW\\/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UOâ pellet fuel. Two of the fuel rods are

  17. Computational Assessment of Burnup-Dependent Fuel Failure Thresholds for Reactivity Initiated Accidents

    Microsoft Academic Search

    Lars Olof JERNKVIST

    2006-01-01

    Best-estimate computational methods are here used to analyse the thermo-mechanical behaviour of high-burnup UO2 fuel rods under postulated reactivity initiated accidents in light water reactors. The considered accident scenarios are the hot zero power rod ejection accident in pressurised water reactors and the cold zero power control rod drop accident in boiling water reactors. For these accidents, fuel enthalpy thresholds

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

    SciTech Connect

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

    2011-01-01

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

  19. Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup

    SciTech Connect

    Chung, H.M.; Kassner, T.F.

    1996-12-01

    High-burnup fuel failure during a reactivity-initiated accident has been the subject of safety-related concern. Because of wide variations in metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior from major tests in NSRR and CABRI reactors. In this paper, a failure model based on fracture toughness and microstructural characteristics is proposed in which fracture toughness of high-burnup cladding is assumed to be sensitive to temperature and exhibit ductile-brittle transition phenomena similar to those of irradiated bcc alloys. Significant effects of temperature and shape of the pulse are predicted when a simulated test is conducted near the material`s transition temperature. Temperature dependence of fracture toughness is, in turn, sensitive to cladding microstructure such as density, distribution, and orientation of hydrides, oxygen distribution in the metallic phase, and irradiation-induced damage. Because all these factors are strongly influenced by corrosion, the key parameters that influence susceptibility to failure are oxide layer thickness and hydriding behavior. Therefore, fuel failure is predicted to be strongly dependent on cladding axial location as well as on burnup. 10 figs, 21 refs.

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

    NASA Astrophysics Data System (ADS)

    Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

    2014-06-01

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

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

    SciTech Connect

    Oggianu, Stella Maris; No, Hee Cheon; Kazimi, Mujid S. [Massachusetts Institute of Technology (United States)

    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.

  2. Nuclide Composition Benchmark Data Set for Verifying Burnup Codes on Spent Light Water Reactor Fuels

    SciTech Connect

    Nakahara, Yoshinori; Suyama, Kenya; Inagawa, Jun; Nagaishi, Ryuji; Kurosawa, Setsumi; Kohno, Nobuaki; Onuki, Mamoru; Mochizuki, Hiroki [Japan Atomic Energy Research Institute (Japan)

    2002-02-15

    To establish a nuclide composition benchmark data set for the verification of burnup codes, destructive analyses of light water reactor spent-fuel samples, which were cut out from several heights of spent-fuel rods, were carried out at the analytical laboratory at the Japan Atomic Energy Research Institute. The 16 samples from three kinds of pressurized water reactor (PWR) fuel rods and the 18 samples from two boiling water reactor (BWR) fuel rods were examined. Their initial {sup 235}U enrichments and burnups were from 2.6 to 4.1% and from 4 to 50 GWd/t, respectively. One PWR fuel rod and one BWR fuel rod contained gadolinia as a burnable poison. The measurements for more than 40 nuclides of uranium, transuranium, and fission product elements were performed by destructive analysis using mass spectrometry, and alpha-ray and gamma-ray spectrometry. Burnup for each sample was determined by the {sup 148}Nd method. The analytical methods and the results as well as the related irradiation condition data are compiled as a complete benchmark data set.

  3. Advanced MOX Core Design Study of Sodium Cooled Reactors in Current Feasibility Study on Commercialized Fast Reactor Cycle Systems in Japan

    SciTech Connect

    Mizuno, T.; Niwa, H. [Japan Nuclear Cycle development institute, O-arai Engineering Center, 4002 Narita-cho, O-arai-machi, Higashi-ibaraki-gun, Ibaraki-ken, 311-1393 (Japan)

    2002-07-01

    The Sodium cooled MOX core design studies are performed with the target burnup of 150 GWd/t and measures against the recriticality issues in core disruptive accidents (CDAs). Four types of core are comparatively studied in viewpoints of core performance and reliability. Result shows that all the types of core satisfy the target and that the homogeneous core with axial blanket partial elimination subassembly is the most superior concept in case the effectiveness of measures against recriticality issues by the axial blanket partial elimination is assured. (authors)

  4. 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 and predominantly scattering isotopes. When the concentration of resonant isotopes is small, its presence does not affect the flux shape which is smooth. But when the concentration becomes high, there will be dips in the flux where resonances of the isotopes occur. This will affect the reaction rate, which is a product of cross section and flux. The reaction rate will thus be lower than that when one does not consider the flux dip. This is the phenomenon of self shielding. Self shielding treatment is thus a very important aspect of reactor lattice analysis code. This needs to be correctly modelled to obtain a physically sound and acceptable solution. In this research we will be looking into behaviour of the advanced self shielding models that have been incorporated in the code DRAGON Version4. The self shielding models are primarily classified into two broad groups, which are based on "equivalence in dilution" and "subgroup approach". These self shielding models will be tested against a variety of lattices which include Canada Deuterium Uranium (CANDU-6), CANDU-New Generation (CANDU-NG), Light Water Reactor (LWR), and High Conversion Light Water Reactor (HCLWR). The fuel composition will vary from natural uranium oxide to enriched uranium oxide and plutonium-uranium mixed oxide (MOX). We will also consider the presence of strong neutron absorbers like gadolinium and dysprosium in the lattice. The coolant/moderator chosen for the analysis will be light water/heavy water or a combination. The lattice geometry will vary from square, hexagonal and annular. Thus a broad spectrum of lattices will be analysed to assess the behaviour of advanced self shielding models. The results obtained using DRAGON will be validated against that obtained using Monte Carlo code MCNP5. The reference solutions for all situations will be provided by MCNP5. The depletion behaviour of any lattice will depend on the power or flux normalization that is considered. In general the flux in various regions is estimated with reference to a single neutron absorbed a

  5. Core design and safety studies for a small modular fast reactor

    SciTech Connect

    Yang, W. S.; Cahalan, J. E.; Dunn, F. E. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2006-07-01

    The paper describes the core design and performance characteristics and the safety analysis results for a 50 MWe small modular fast reactor design that was developed jointly by ANL, CEA, and JNC as an international collaborative effort. The main goal in the core design was to achieve a 30-year lifetime with no refueling. In order to minimize the burnup reactivity swing, metal fuel with a high heavy metal volume fraction was selected. To enhance the proliferation resistance and actinide transmutation, all the transuranic (TRU) elements recovered from light water reactor spent fuel were used in a ternary alloy form of U-TRU-10Zr. A 125 MWt core design was developed, for which the burnup reactivity swing was only 1.6$ over the 30-year core lifetime. The average discharge burnup was 87 MWd/kg, and the maximum sodium void worth was 4.65$. The evaluated reactivity coefficients provided sufficient negative feedbacks. Shutdown margins of control systems were confirmed. Steady-state thermal-hydraulic analysis results showed that peak 2{sigma} cladding inner-wall and fuel centerline temperatures were less than design limits with sufficient margins. Detailed transient analyses for the total loss of power to reactor and intermediate coolant pumps showed that no fuel damage or cladding failure would occur, even when multiple safety systems were assumed to malfunction. (authors)

  6. Automated Design and Optimization of Pebble-bed Reactor Cores

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2010-07-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  7. Viscosity of the Earth's Core

    Microsoft Academic Search

    Roger F. Gans

    1972-01-01

    The viscosity of the earth's core is probably the least well-known physical property of the earth. Miki [1952] gives an estimate, based on a theoretical calculation, that the dynamic viscosity lies between 10 - and 10 - poise. Malkus [1968] suggests the range 10 -' to 1 poise. Attenuation of S waves reflected from the core [Sato and Espinosa, 1967b;

  8. MCNP Simulation of Void Reactivity in a Simplified CANDU Core Sub-region

    NASA Astrophysics Data System (ADS)

    Rahnema, F.; Mosher, S.; Pitts, M.; Akhtar, P.; Serghiuta, D.

    The Monte Carlo code MCNP with a continuous-energy ENDF/B-VI cross section library at the hot operating condition was used to determine the impact of the core environment on void reactivity in a sub-region of a simplified CANDU-6 core of 4 x 3 x 6 cell-size. The net (combined) impact of the adjuster rods, axial leakage and cell-to-cell radial leakage (due to fuel burnup variation in the core) was estimated to be between 1.44 ± 0.37 and 1.96 ± 0.39 mk (10-3k).

  9. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    NASA Astrophysics Data System (ADS)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  10. W isotope constraints on the mechanisms of terrestrial core formation

    Microsoft Academic Search

    T. Kleine; B. Bourdon

    2006-01-01

    The 182Hf-182W system should, in principle, be well suited to date core formation in Earth but the calculated ages strongly depend on the core formation\\/accretion model applied [e.g.,1-5]. An essential parameter for calculating Hf-W ages of core formation is the degree of metal-silicate equilibration during core formation. For instance, early core formation with complete metal-silicate equilibration and late core formation

  11. Dislocation core energies and core fields from first principles.

    PubMed

    Clouet, Emmanuel; Ventelon, Lisa; Willaime, F

    2009-02-01

    Ab initio calculations in bcc iron show that a 111 screw dislocation induces a short-range dilatation field in addition to the Volterra elastic field. This core field is modeled in anisotropic elastic theory using force dipoles. The elastic modeling thus better reproduces the atom displacements observed in ab initio calculations. Including this core field in the computation of the elastic energy allows deriving a core energy which converges faster with the cell size, thus leading to a result which does not depend on the geometry of the dislocation array used for the simulation. PMID:19257518

  12. Application of perturbation theory to lattice calculations based on method of cyclic characteristics

    NASA Astrophysics Data System (ADS)

    Assawaroongruengchot, Monchai

    Perturbation theory is a technique used for the estimation of changes in performance functionals, such as linear reaction rate ratio and eigenvalue affected by small variations in reactor core compositions. Here the algorithm of perturbation theory is developed for the multigroup integral neutron transport problems in 2D fuel assemblies with isotropic scattering. The integral transport equation is used in the perturbative formulation because it represents the interconnecting neutronic systems of the lattice assemblies via the tracking lines. When the integral neutron transport equation is used in the formulation, one needs to solve the resulting integral transport equations for the flux importance and generalized flux importance functions. The relationship between the generalized flux importance and generalized source importance functions is defined in order to transform the generalized flux importance transport equations into the integro-differential equations for the generalized adjoints. Next we develop the adjoint and generalized adjoint transport solution algorithms based on the method of cyclic characteristics (MOCC) in DRAGON code. In the MOCC method, the adjoint characteristics equations associated with a cyclic tracking line are formulated in such a way that a closed form for the adjoint angular function can be obtained. The MOCC method then requires only one cycle of scanning over the cyclic tracking lines in each spatial iteration. We also show that the source importance function by CP method is mathematically equivalent to the adjoint function by MOCC method. In order to speed up the MOCC solution algorithm, a group-reduction and group-splitting techniques based on the structure of the adjoint scattering matrix are implemented. A combined forward flux/adjoint function iteration scheme, based on the group-splitting technique and the common use of a large number of variables storing tracking-line data and exponential values, is proposed to reduce the computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR-BOC, CVR-EOC and keff-EOC adjustment of a CANDU lattice of which the burnup period is extended f

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

    SciTech Connect

    Cheatham, Jesse R [ORNL] [ORNL; Francis, Matthew W [ORNL] [ORNL

    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.

  14. Analysis of high burnup spent nuclear fuel by ICP-MS

    Microsoft Academic Search

    S. F. Wolf; D. L. Bowers; J. C. Cunnane

    2005-01-01

    Summary  We have used inductively coupled plasma mass spectrometry (ICP-MS) as the primary tool for determining concentrations of a suite of nuclides in samples excised from high-burnup spent nuclear fuel rods taken from light water nuclear reactors. The complete analysis included the determination of 95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 137Cs, 143Nd, 145Nd,148Nd,147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 151Eu, 153Eu, 155Eu, 155Gd, 237Np,

  15. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP

    SciTech Connect

    Primm, Trent [ORNL] [ORNL; Chandler, David [ORNL] [ORNL

    2009-01-01

    Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

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

    PubMed

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

    2012-06-01

    The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium. PMID:22476019

  17. Sodium fast reactor evaluation: Core materials

    NASA Astrophysics Data System (ADS)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

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

    SciTech Connect

    Teague, Melissa C. [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Gorman, Brian P. [Colorado School of Mines, Golden, CO (United States); Miller, Brandon D. [Idaho National Laboratory (INL), Idaho Falls, ID (United States); King, Jeffrey [Colorado School of Mines, Golden, CO (United States)

    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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-01-01

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

  1. Thermal Diffusivity Of Homogeneous SBR MOX Fuel With A Burn-up of 35 MWd/kgHM

    SciTech Connect

    Staicu, D.; Pagliosa, G.; Papaioannou, D.; Rondinella, V.V.; Cozzo, C.; Konings, R.; Walker, C.T. [European Commission, Joint Research Centre, Institute for Transuranium Element, P.O. Box 2340, D-76125 Karlsruhe (Germany); Barker, M. [Nexia Solutions, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom); Weston, R. [British Nuclear Group Sellafield Ltd, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom)

    2007-07-01

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

  2. Pre-supernova neutrino emissions from ONe cores in the progenitors of core-collapse supernovae: are they distinguishable from those of Fe cores?

    E-print Network

    Kato, Chinami; Yamada, Shoichi; Takahashi, Koh; Umeda, Hideyuki; Yoshida, Takashi; Ishidoshiro, Koji

    2015-01-01

    Aiming to distinguish two types of progenitors of core collapse supernovae, i.e., one with a core composed mainly of oxygen and neon (abbreviated as ONe core) and the other with an iron core (or Fe core), we calculated the luminosities and spectra of neutrinos emitted from these cores prior to gravitational collapse, taking neutrino oscillation into account. We found that the total energies emitted as $\\bar{\

  3. F-layer formation in the outer core with asymmetric inner core growth

    NASA Astrophysics Data System (ADS)

    Deguen, Renaud; Olson, Peter; Reynolds, Evan

    2014-05-01

    Numerical calculations of thermochemical convection in a rotating, electrically conducting fluid sphere with heterogeneous boundary conditions are used to model effects of asymmetric inner core growth. With heterogeneous inner core growth but no melting, outer core flow consists of intense convection where inner core buoyancy release is high, weak convection where inner core buoyancy release is low, and large scale, mostly westward flow in the form of spiraling gyres. With localized inner core melting, outer core flow includes a gravity current of dense fluid that spreads over the inner core boundary, analogous to the seismic F-layer. An analytical model for gravity currents on a sphere connects the structure of the dense layer to the distribution of inner core melting and solidification. Predictions for F-layer formation by asymmetric inner core growth include large-scale asymmetric gyres below the core-mantle boundary and eccentricity of the geomagnetic field.

  4. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail

    2010-12-01

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

  5. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect

    Permana, Sidik [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Sekimoto, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Waris, Abdul; Subhki, Muhamad Nurul [Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Ismail, [BAPETEN (Indonesia)

    2010-12-23

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

  6. Calculation of radiation induced swelling of uranium mononitride using the digital computer program CYGRO 2

    NASA Technical Reports Server (NTRS)

    Davison, H. W.; Fiero, I. B.

    1971-01-01

    Fuel volume swelling and clad diametral creep strains were calculated for five fuel pins, clad with either T-111 (Ta-8W-2.4Hf) or PWC-11 (Nb-1Zr-0.1C). The fuel pins were irradiated to burnups between 2.7 and 4.6%. Clad temperatures were between 1750 and 2400 F (1228 and 1589 K). The maximum percentage difference between calculated and experimentally measured values of volumetric fuel swelling is 60%.

  7. An object-oriented implementation of the invariant imbedding method in reactor physics calculations

    NASA Astrophysics Data System (ADS)

    Forsbacka, Matthew John

    The method of imbedded invariants was applied to reactor physics problems using an object-oriented computational approach. To my knowledge this is the first application of object-oriented methods to such problems. In so doing, a three-dimensional implementation of invariant imbedding was realized for the first time in a practical reactor physics problem. Invariant imbedding presents an alternative approach to solving the traditional reactor physics problems of calculating the neutron multiplication factor, ksb{eff}, and the neutron distribution in a nuclear system. In essence, invariant imbedding reformulates a traditional boundary value problem into the form of an initial value problem where related subspaces of a problem are linked together under a common imbedding parameter. Using this mathematical approach, the neutronic characteristics of elements comprising a neutron multiplying system are coupled together to recursively calculate ksb{eff} for any system arrangement. System mesh elements are characterized using a Monte Carlo code, so that each element can contain any unique interior geometry and/or material composition. Cells do not have to have the same shapes as long as their bounding faces match. Thus, three-dimension criticality calculations are possible for any system change ranging from fuel burnup to severe core-disruption accident consequence analyses. The accuracy of the method was demonstrated by analysing a set of critical experiments done at Oak Ridge National Laboratory. A three-dimensional model was demonstrated to converge to a benchmark solution as the mesh element size was reduced. A correction factor was implemented to speed up convergence for large meshes.

  8. The BURNUP package of applied programs used for computing the isotopic composition of materials of an operating nuclear reactor

    SciTech Connect

    Yudkevich, M. S., E-mail: umark@adis.vver.kiae.ru [National Research Centre Kurchatov Institute (Russian Federation)

    2012-12-15

    This paper described the procedure of implementation and the possibilities of the BURNUP program. The purpose of the program is to predict the change in the nuclear composition of the materials of which a reactor is made in the course of its run and compute the radiation characteristics of the materials after their irradiation.

  9. Fission gas and iodine release measured in IFA430 up to 15 GWd\\/t UOâ burnup. [PWR; BWR

    Microsoft Academic Search

    A. D. Appelhans; J. A. Turnbull; R. J. White

    1983-01-01

    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper

  10. Boiling water reactor fuel behavior at burnup of 26 GWd\\/tonne U under reactivity-initiated accident conditions

    Microsoft Academic Search

    Takehiko Nakamura; Makio Yoshinaga; Makoto Sobajima; Kiyomi Ishijima; Toshio Fujishiro

    1994-01-01

    Irradiated boiling water reactor (BWR) fuel behavior under reactivity-initiated accident (RIA) conditions was investigated in the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute. Short test fuel rods, refabricated from a commercial 7 x 7 type BWR fuel rod at a burnup of 26 GWd\\/ tonne U, were pulse irradiated in the NSRR under simulated cooled

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

    Microsoft Academic Search

    Michael Salay; Randall O. Gauntt; Richard Y. Lee; Dana Auburn Powers; Mark Thomas Leonard

    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

  12. Air core pulse transformer design

    Microsoft Academic Search

    J. P. O'Loughlin; J. D. Sidler; Gerry J. Rohwein

    1988-01-01

    Cylindrical-air-core pulse transformers capable of passing high-voltage\\/high-energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters than iron or ferrite core units. Special computer codes were written to evaluate their performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation

  13. Viscosity of the earth's core.

    NASA Technical Reports Server (NTRS)

    Gans, R. F.

    1972-01-01

    Calculation of the viscosity of the core at the boundary of the inner and outer core. It is assumed that this boundary is a melting transition and the viscosity limits of the Andrade (1934,1952) hypothesis (3.7 to 18.5 cp) are adopted. The corresponding kinematic viscosities are such that the precessional system explored by Malkus (1968) would be unstable. Whether it would be sufficiently unstable to overcome a severely subadiabatic temperature gradient cannot be determined.

  14. Realizing low loss air core photonic crystal fibers by exploiting an antiresonant core surround

    Microsoft Academic Search

    P. J. Roberts; D. P. Williams; B. J. Mangan; H. Sabert; F. Couny; W. J. Wadsworth; T. A. Birks; J. C. Knight; P. St. J. Russell

    2005-01-01

    The modal properties of an air core photonic crystal fiber which incorporates an anti-resonant feature within the region that marks the transition between the air core and the crystal cladding are numerically calculated. The field intensity at the glass\\/air interfaces is shown to be reduced by a factor of approximately three compared to a fiber with more conventional core surround

  15. Ris Report No. 286 J-Danish Atomic Energy Commission

    E-print Network

    Spheres 20 4 . 3 . Methods of Calculation 22 4.4. Results of Sphere Calculations 24 5. Unit Cell Burn-up Calculations for the Yankee Rowe Reactor.. 31 5.1. The Yankee Core and the EYC Project 32 5.2. Unit Cell Burn calculations for spherical bench- mark experiments, burn-up calculations for the Yankee Rowe reactor

  16. Oxygen stoichiometry shift of irradiated LWR-fuels at high burn-ups: Review of data and alternative interpretation of recently published results

    NASA Astrophysics Data System (ADS)

    Spino, J.; Peerani, P.

    2008-03-01

    The available oxygen potential data of LWR-fuels by the EFM-method have been reviewed and compared with thermodynamic data of equivalent simulated fuels and mixed oxide systems, combined with the analysis of lattice parameter data. Up to burn-ups of 70-80 GWd/tM the comparison confirmed traditional predictions anticipating the fuels to remain quasi stoichiometric along irradiation. However, recent predictions of a fuel with average burn-up around 100 GWd/tM becoming definitely hypostoichiometric were not confirmed. At average burn-ups around 80 GWd/tM and above, it is shown that the fuels tend to acquire progressively slightly hyperstoichiometric O/M ratios. The maximum derived O/M ratio for an average burn-up of 100 GWd/tM lies around 2.001 and 2.002. Though slight, the stoichiometry shift may have a measurable accelerating impact on fission gas diffusion and release.

  17. 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 [Nuclear Physics and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Ferhat, A. [National Nuclear Energ Agency of Indonesia (BATAN) (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

    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.

  18. Calculator Cookery.

    ERIC Educational Resources Information Center

    Humphreys, Casey; And Others

    This valuable collection of materials was developed to incorporate the calculator as an instructional aid in ninth- and tenth-grade general and basic mathematics classes. The materials are also appropriate for grades 7 and 8. After an introductory section which teaches the use of the calculator, four games and activities are described. For these…

  19. Phasor Calculator

    NSDL National Science Digital Library

    Reeder, Nick

    Nick Reeder from Sinclair Community College created this interactive tool to compute with phasors when you're analyzing AC circuits. The phasor calculator allows you to add, subtract, multiply, or divide phasors. Use E to enter numbers in exponential form, such as 50E3 for 50 kilo. Press the = button to calculate result. Use the 1/x buttons to invert either operand.

  20. BTU Calculator

    NSDL National Science Digital Library

    A calculator that estimates the heating needs of a room, a combination of rooms, or an entire home. Enter the length and width of the area to be heated and select the climate and insulation factors from the pop-up boxes. A Java version of this calculator is also available.

  1. Benchmark data for validating irradiated fuel compositions used in criticality calculations

    SciTech Connect

    Bierman, S.R.; Talbert, R.J.

    1994-10-01

    To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays have been obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of a Pressurized Water Reactor fuel rod and represent radiation exposures of about 37, 27, and 44 GWd/MTU. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input.

  2. On the thermal conductivity of UO 2 nuclear fuel at a high burn-up of around 100 MWd\\/kgHM

    Microsoft Academic Search

    C. T. Walker; D. Staicu; M. Sheindlin; D. Papaioannou; W. Goll; F. Sontheimer

    2006-01-01

    A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102MWd\\/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was

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

    Microsoft Academic Search

    Tomoyuki Uwaba; Masahiro Ito; Koji Maeda

    2011-01-01

    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 130GWd\\/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,

  4. Fission product release and microstructure changes during laboratory annealing of a very high burn-up fuel specimen

    Microsoft Academic Search

    J.-P. Hiernaut; T. Wiss; J.-Y. Colle; H. Thiele; C. T. Walker; W. Goll; R. J. M. Konings

    2008-01-01

    A commercial PWR fuel sample with a local burn-up of about 240MWd\\/kgHM was annealed in a Knudsen cell mass spectrometer system with a heating rate of 10K\\/min up to 2750K at which temperature the sample was completely vaporized. The release of fission gases and fission products was studied as a function of temperature. In one of the runs the heating

  5. Identification of Radial Position of Fission Gas Release in High-Burnup Fuel Pellets under RIA Conditions

    Microsoft Academic Search

    Hideo SASAJIMA; Tomoyuki SUGIYAMA; Toshinori CHUTO; Fumihisa NAGASE; Takehiko NAKAMURA; Toyoshi FUKETA

    2010-01-01

    The radial positions of fission gas release (FGR) in high-burnup fuel pellets were examined after pulseirradiations that simulated reactivity-initiated accident (RIA) conditions in the Nuclear Safety Research Reactor (NSRR). The molar ratio of xenon (Xe) to krypton (Kr) (Xe\\/Kr ratio) in the released gas showed that fission gas was released from the entire region of the pellets of the examined

  6. Isotope correlations for determining the isotopic composition of plutonium in high burnup pressurized water reactor (PWR) samples

    Microsoft Academic Search

    Kihsoo Joe; Young-Shin Jeon; Byung-Chul Song; Sun-Ho Han; Euo-Chang Jung; Kyuseok Song

    2010-01-01

    Correlations among the alpha activity ratios of 238Pu\\/(239Pu+240Pu), the alpha specific activities of Pu and the atom % abundances of Pu isotopes were derived for the plutonium samples obtained from high burnup fuel samples from pressurized water reactors. Using the alpha activity ratios of 238Pu\\/(239Pu+240Pu) determined by alpha spectrometry, the alpha specific activities of Pu as well as the atom

  7. Thermal Diffusivity Of Homogeneous SBR MOX Fuel With A Burn-up of 35 MWd\\/kgHM

    Microsoft Academic Search

    D. Staicu; G. Pagliosa; D. Papaioannou; V. V. Rondinella; C. Cozzo; R. Konings; C. T. Walker; M. Barker; R. Weston

    2007-01-01

    New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd\\/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded 'laser-flash' device and show that the thermal diffusivity increases from the pellet periphery to the centre. Comparison shows that the thermal conductivity is in the same range than

  8. Effect of burn-up on the thermal conductivity of uranium dioxide up to 100.000 MWd t ?1

    Microsoft Academic Search

    C. Ronchi; M. Sheindlin; D. Staicu; M. Kinoshita

    2004-01-01

    The thermal diffusivity and specific heat of reactor-irradiated UO2 fuel have been measured. Starting from end-of-life conditions at various burn-ups, measurements under thermal annealing cycles were performed in order to investigate the recovery of the thermal conductivity as a function of temperature. The separate effects of soluble fission products, of fission gas frozen in dynamical solution and of radiation damage

  9. Improved gas core propulsion model

    SciTech Connect

    Tanner, J.E.

    1993-10-01

    A thermodynamic, radiation transport model of a gas core nuclear propulsion reactor has been developed in one-dimensional, spherical geometry, which satisfies local energy balance and allows for arbitrary variation of fuel/propellant ratio and flow rate as functions of radius. Initial cases calculated yield specific impulses of about 1150 sec, but very low thrusts ranging 5--10 kN.

  10. Calculating Flows In Turbomachine Channels

    NASA Technical Reports Server (NTRS)

    Schumann, Lawrence F.

    1989-01-01

    Noniterative integral-entrainment method yields good approximations. Method of approximate calculation of flow in channel of turbomachine based on interaction of viscous flow in boundary layers with inviscid flow in core layer. Faster and more robust than other approximate methods of same type. Suitable for use in preliminary calculations for design and for off-design operation of turbomachinery. Flows in conical diffuser channels represented by two-dimensional boundary-layer and one-dimensional core flows described by equations of new method.

  11. Improvements in boiling water reactor uranium utilization and operating experience with burnup increase

    Microsoft Academic Search

    Takaaki Mochida; Katsumasa Haikawa; Junichi Yamashita; Akira Nishimura; Yutaka Iwata; Shiroh Arai

    1996-01-01

    A boiling water reactor (BWR) core design for better uranium utilization is presented, and its validity is demonstrated through simulation and operation data. Together with the axial power flattening obtained by an axially zoned enrichment core, uranium utilization improvement techniques such as an axial blanket for neutron leakage reduction, a low leakage loading pattern, an improved local enrichment distribution in

  12. 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.

  13. Microchemical study of high-burnup CANDU ® fuel by imaging-XPS

    NASA Astrophysics Data System (ADS)

    Do, Than; Irving, Karen G.; Hocking, William H.

    2008-12-01

    An advanced facility for characterization of highly radioactive materials by Imaging X-ray Photoelectron Spectroscopy (XPS) has been developed at the Chalk River Laboratories (CRL), based upon over a decade of prior experience with a prototype system. Auxiliary electron and ion guns provide additional in situ capabilities for scanning electron microscopy (SEM), scanning Auger microscopy (SAM) and composition depth profiling. The application of this facility to the characterization of irradiated fuel materials will be illustrated with selected results taken from a detailed study of the microchemistry at the fuel-sheath interface in a CANDU fuel element that was irradiated to extended burnup in the NRU (National Research Universal) reactor at CRL. Inside surfaces of the end caps and the welds between the sheath and the end caps as well as the thin-walled Zircaloy-4 sheath were analyzed. The in situ SEM capability was essential for selecting different areas on each sample, such as sheath locations with and without a visible retained CANLUB graphite layer, for XPS analysis. Effective infiltration of segregated fission products, especially cesium, into the graphite was demonstrated by depth profiling. A richer chemistry of segregated fission products was found on the end caps than on the sheath with elevated levels of barium, strontium, tellurium, iodine and cadmium as well as cesium. The results are consistent with current understanding of the primary migration route for fission products to the sheath and also indicate that the CANLUB layer functions as a chemical rather than a physical barrier to segregated fission products.

  14. MEMS Calculator

    National Institute of Standards and Technology Data Gateway

    SRD 166 MEMS Calculator (Web, free access)   This MEMS Calculator determines the following thin film properties from data taken with an optical interferometer or comparable instrument: a) residual strain from fixed-fixed beams, b) strain gradient from cantilevers, c) step heights or thicknesses from step-height test structures, and d) in-plane lengths or deflections. Then, residual stress and stress gradient calculations can be made after an optical vibrometer or comparable instrument is used to obtain Young's modulus from resonating cantilevers or fixed-fixed beams. In addition, wafer bond strength is determined from micro-chevron test structures using a material test machine.

  15. Plasma core reactor applications

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Rodgers, R. J.

    1976-01-01

    Analytical and experimental investigations were conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration. Axial working fluid channels are located along a fraction of each cavity peripheral wall. Results of calculations for outward-directed radiant energy fluxes corresponding to radiating temperatures of 2000 to 5000 K indicate total operating pressures from 80 to 650 atm, centerline temperatures from 6900 to 30,000 K, and total radiated powers from 25 to 2500 MW, respectively. Applications are described for this type of reactor such as (1) high-thrust, high specific impulse space propulsion, (2) highly efficient systems for generation of electricity, and (3) hydrogen or synthetic fuel production systems using the intense radiant energy fluxes.

  16. Evolution of First Cores and Formation of Stellar Cores in Rotating Molecular Cloud Cores

    NASA Astrophysics Data System (ADS)

    Saigo, Kazuya; Tomisaka, Kohji; Matsumoto, Tomoaki

    2008-02-01

    We followed the collapse of cloud cores with various rotation speed and density frustrations using three-dimensional hydrodynamical simulations by assuming a barotropic equation of state and examined the comprehensive evolution paths from the rotation molecule cloud core to stellar core. We found that the evolutionary paths depend only on the angular velocity of initial cloud core ?c0. These evolutionary paths agree well with predictions of Saigo and Tomisaka's quasi-equilibrium axisymmetric models and SPH calculations of Bate. Evolutionary paths are qualitatively classified into three types. (1) A slowly rotating cloud with ?c0<0.01/tff=0.05(?c0/10-19 g cm-3)1/2 rad Myr-1 shows spherical-type evolution, where ?c0 is the initial central density. Such a cloud forms a first core which is mainly supported by the thermal pressure. The first core has a small mass of Mcore~0.01 Msolar and a short lifetime of a few ×100 yr. After exceeding the H2 dissociation density ?~=5.6×10-8 g cm-3, it begins the second collapse, and the whole of the first core accretes onto the stellar core/disk within a few free-fall timescales. (2) A rotating cloud with 0.01/tffcore becomes a centrifugally supported massive disk with Mcore~a few×0.01-0.1 Msolar and the lifetime is a few thousand years. The first core is unstable against nonaxisymmetric dynamic instability and forms spiral arms. The gravitational torque through spiral structure extracts angular momentum from the central region to the outer region of the first core. And only a central part with r~1 AU begins the second collapse after exceeding dissociation density. However, the outer remnant disk keeps its centrifugal balance after stellar core formation. It seems that this remnant of the first core should control the mass and angular momentum accretion onto the newborn stellar system. (3) A rotating cloud with 0.05/tff<~?c0 tends to fragment into binary or multiple during the first core phase.

  17. A FEASIBILITY AND OPTIMIZATION STUDY TO DETERMINE COOLING TIME AND BURNUP OF ADVANCED TEST REACTOR FUELS USING A NONDESTRUCTIVE TECHNIQUE

    SciTech Connect

    Jorge Navarro

    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 for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

  18. 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 [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States)] [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States); Wagner, John C. [Oak Ridge National Laboratory (United States)] [Oak Ridge National Laboratory (United States)

    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 fuel storage and transportation. This paper discusses the staff's preliminary considerations on the safety implication of fuel reconfiguration with respect to nuclear safety (subcriticality control), radiation shielding, containment, the performance of the thermal functions of the packages, and the retrievability of the contents from regulatory perspective. (authors)

  19. Inner Core Rotation from Geomagnetic Westward Drift and a Stationary Spherical Vortex in Earth's Core

    NASA Technical Reports Server (NTRS)

    Voorhies, Coerte V.

    1998-01-01

    The idea that geomagnetic westward drift indicates convective leveling of the planetary momentum gradient within Earth's core is pursued in search of a differentially rotating mean state, upon which various oscillations and secular effects might be superimposed. The desired state conforms to roughly spherical boundary conditions, minimizes dissipative interference with convective cooling in the bulk of the core, yet may aid core cooling by depositing heat in the uppermost core and lower mantle. The variational calculus of stationary dissipation applied to a spherical vortex within the core yields an interesting differential rotation profile, akin to spherical Couette flow bounded by thin Hartmann layers. Four boundary conditions are required. To concentrate shear induced dissipation near the core-mantle boundary, these are taken to be: (i) no-slip at the core-mantle interface; (ii) geomagnetically estimated bulk westward flow at the base of the core-mantle boundary layer; (iii) no-slip at the inner-outer core interface; and, to describe magnetic locking of the inner core to the deep outer core; (iv) hydrodynamically stress-free at the inner-outer core boundary. By boldly assuming the axial core angular momentum anomaly to be zero, the super-rotation of the inner core relative to the mantle is calculated to be at most 1.5 deg./yr.

  20. Start-up fuel and power flattening of sodium-cooled candle core

    SciTech Connect

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, Hiroshi [University of California, Berkeley, CA 94720 (United States)

    2013-07-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing.

  1. Inflation Calculator

    NSDL National Science Digital Library

    Friedman, S. Morgan.

    This simple inflation calculator uses the Consumer Price Index to adjust any given amount of money, from 1800 to 1998. Creator S. Morgan Friedman uses data from the Historical Statistics of the United States for statistics predating 1975 and the annual Statistics Abstracts of the United States for data from 1975 to 1998. Links to other online inflation information are also included.

  2. Calculation Nation

    NSDL National Science Digital Library

    2011-01-01

    This web site, which is part of the NCTM Illuminations project, allows students to challenge themselves or opponents from anywhere in the world by playing games that are organized around content from the upper elementary and middle grades math curriculum. The games allow students to learn about fractions, factors, multiples, symmetry, as well as practice important skills like basic multiplication and calculating area.

  3. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Microsoft Academic Search

    F. Faghihi; S. M. Mirvakili

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (?ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and

  4. Benchmark Calculations for Standard and DUPIC CANDU Fuel Lattices Compared with the MCNP-4B Code

    SciTech Connect

    Roh, Gyuhong; Choi, Hangbok [Korea Atomic Energy Research Institute (Korea, Republic of)

    2000-10-15

    Cell-code benchmark calculations have been performed for the standard CANDU and DUPIC CANDU fuel lattices compared with the MCNP-4B code. To consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The lattice codes WIMS-AECL and HELIOS were then benchmarked by the MCNP code for the major physics parameters such as burnup reactivity, coolant void reactivity, fuel temperature coefficient, etc. The calculations have shown that the physics parameters estimated by the lattice codes are consistent with those by MCNP. However, there is a tendency that the error increases slightly when the fuel burnup is high. This study has shown that the WIMS-AECL produces reliable results for CANDU fuel analysis. However, it is recommended that the cross-section library be updated to be used for the high-burnup fuels even though the current results are generally acceptable. This study has also shown that the HELIOS code has the potential to be used for CANDU fuel lattice analysis in the future.

  5. Calculation Software

    NASA Technical Reports Server (NTRS)

    1994-01-01

    MathSoft Plus 5.0 is a calculation software package for electrical engineers and computer scientists who need advanced math functionality. It incorporates SmartMath, an expert system that determines a strategy for solving difficult mathematical problems. SmartMath was the result of the integration into Mathcad of CLIPS, a NASA-developed shell for creating expert systems. By using CLIPS, MathSoft, Inc. was able to save the time and money involved in writing the original program.

  6. Broken Calculator

    NSDL National Science Digital Library

    Mandy Barrow

    2008-01-01

    This interactive applet helps students develop fluency and flexibility with numbers. At each of 6 difficulty levels the user is presented with 8 target numbers and a partial set of keys on a basic calculator (does not follow order of operations). The goal is to use the given keys to make as many of the target numbers as possible within the 3-minute time limit. Some levels include memory keys.

  7. Radiation Damage Calculations for the FUBR and BEATRIX Irradiations of Lithium Compounds in EBR-II and FFTF

    SciTech Connect

    LR Greenwood

    1999-06-17

    The Fusion Breeder Reactor (FUBR) and Breeder Exchange Matrix (BEATRIX) experiments were cooperative efforts by members of the International Energy Agency to investigate the irradiation behavior of solid breeder materials for tritium production to support future fusion reactors. Lithium ceramic materials including Li{sub 2}O, LiAlO{sub 2}, Li{sub 4}SiO{sub 4}, and Li{sub 2}ZrO{sub 3} with varying {sup 6}Li enrichments from 0 to 95% were irradiated in a series of experiments in the Experimental Breeder Reactor (EBR II) and in the Fast Flux Test Facility (FFTF) over a period of about 10 years from 1982 to 1992. These experiments were characterized in terms of the nominal fast neutron fluences and measured {sup 6}Li burnup factors, as determined by either mass spectrometry or helium measurements. Radiation damage in these compounds is caused by both the {sup 6}Li-burnup reaction and by all other possible neutron reactions with the atoms in the compound materials. In this report, displacements per atom (dpa) values have been calculated for each type of material in each of the various irradiations that were conducted. Values up to 11% {sup 6}Li-burnup and 130 dpa are predicted for the longest irradiations. The dpa cross sections were calculated for each compound using the SPECOMP computer code. Details of the dpa calculations are presented in the report. Total dpa factors were determined with the SPECTER computer code by averaging the dpa cross sections over the measured or calculated neutron flux spectra for each series of irradiations. Using these new calculations, previously measured radiation damage effects in these lithium compounds can be compared or correlated with other irradiation data on the basis of the dpa factor as well as {sup 6}Li-burnup.

  8. Passive Safety Small Reactor for Distributed Energy Supply: Heavy Water Mixing Core

    SciTech Connect

    Ken-ichi Sawada; Naoteru Odano [National Maritime Research Institute, 6-38-1, Shinkawa, Mitaka-shi, Tokyo 181-0004 (Japan); Toshihisa Ishida [Kobe University, Kobe 657-8501 (Japan)

    2006-07-01

    The purpose of this paper is to study the core performance of passive safety small reactor for distributed energy supply by changing the heavy water (D{sub 2}O) concentration in the mixed coolant together with the fuel pitch. The long core life with conditions of the excessive reactivity of 2 %{delta}k/k, the reactivity shutdown margin of 1 %{delta}k/k and the negative coolant temperature reactivity coefficient is attained for the case of D{sub 2}O concentration of 60 % with 10 % enrichment gadolinia (Gd{sub 2}O{sub 3}) doped fuel rods. This D{sub 2}O core has a shorter core life 4.14 years than the original light water (H{sub 2}O) core 4.76 years, while it needs a larger core size. However, changing the D{sub 2}O concentration on the way during the burn-up shows a possibility of extending more the core life than that of the original H{sub 2}O core. (authors)

  9. Rayleigh-Taylor overturn in supernova core collapse

    NASA Astrophysics Data System (ADS)

    Smarr, L.; Wilson, J. R.; Barton, R. T.; Bowers, R. L.

    1981-06-01

    A two dimensional radiation diffusion coupled to hydrodynamics calculation of nonspherical instabilities in the collapsed core from a massive star is performed. The core properties are taken from a one-dimensional collapse calculated with detailed microphysics and neutrino transport. The shocked outer core (mass between 0.7 and 1.3 solar masses) is found to contain three subregions. The innermost is doubly diffusive (neutron fingers) unstable, the center subregion is dynamically unstable, and the outermost is stable. It is found that the unstable part of the outer core overturns in approximately 5 ms without disturbing the inner unshocked core of mass less than 0.7 solar masses. This overturned outer core expands as a piston, creating an outgoing shock wave which may help power envelope ejection. This outer core overturn would seem to be a generic feature of core collapse which has heretofore been neglected.

  10. Air core pulse transformer design

    SciTech Connect

    O'Loughlin, J.P.; Sidler, J.D.; Rohwein, G.J.

    1988-01-01

    Cylindrical air core pulse transformers capable of passing high voltage/energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters that iron or ferrite core units. The two salient advantages of the air core transformer are a much lighter weight and a simplified high voltage insulation system. Special computer codes were written to evaluate the performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation stress, evaluation of the waveforms distortion and energy transfer efficiency. Graphical data are given for the optimization in terms of electrical parameters. The results are in agreement with experimental data. It is concluded that air core transformers are feasible operating at hundreds of kilovolts and lens of kioljoules in the microsecond region with energy transfer efficiencies of 70% to 85%. The insulation stresses required are in the 100 to 300 kV/em range. Effects of the high frequency current distribution in the windings and the use of ''slug'' type ferrite cores are also evaluated.

  11. Air core pulse transformer design

    NASA Astrophysics Data System (ADS)

    Oloughlin, J. P.; Sidler, J. D.; Rohwein, G. J.

    Cylindrical air core pulse transformers capable of passing high voltage/energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters that iron or ferrite core units. The two salient advantages of the air core transformer are a much lighter weight and a simplified high voltage insulation system. Special computer codes were written to evaluate the performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation stress, evaluation of the waveforms distortion and energy transfer efficiency. Graphical data are given for the optimization in terms of electrical parameters. The results are in agreement with experimental data. It is concluded that air core transformers are feasible operating at hundreds of kilovolts and tens of kioljoules in the microsecond region with energy transfer efficiencies of 70 to 85 percent. The insulation stresses required are in the 100 to 300 kV/em range. Effects of the high frequency current distribution in the windings and the use of slug type ferrite cores are also evaluated.

  12. Core ionization energies of carbon-nitrogen molecules and solids

    Microsoft Academic Search

    A. Snis; S. F. Matar; O. Plashkevych; H. Ågren

    1999-01-01

    Core ionization energies have been calculated for various carbon-nitrogen molecules and solids. The systems investigated contain many of the bonding possibilities which presumably arise in carbon nitride thin films prepared under varying conditions. The molecular core ionization energies are calculated by the DeltaSCF self-consistent field method. Several singly, doubly, and triply bonded CxNyHz species have been considered. Core ionization energies

  13. 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 ?c = 0.29. Rim-cavities are expected to remain closed at least up to this limit.

  14. 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.

  15. Occlusion Calculator

    PubMed Central

    Hiremath, Anand; Aluckal, Eby

    2015-01-01

    Start with end in mind’ is a popular cliché in orthodontics. This aptly applies to the therapeutic occlusion the orthodontist strives to achieve. Predicting the post treatment occlusion is an essential component of treatment planning. When no extractions or symmetric extractions are done predicting the final occlusion is somewhat easy. Prediction is challenging when we do unconventional and/or asymmetric extractions. To aid this decision Kesling proposed the ‘Kesling Setup’. Though it serves the purpose acceptably; it is time, energy and money consuming. We have developed a model which can help us visualize the final occlusion in matter of seconds. Although this model is primarily intended for orthodontic postgraduate teaching, it can be of considerable use even to a seasoned orthodontist. The regular use of “Orthodontic Calculator” in our department is a testimony to its usefulness. PMID:25738101

  16. Probability Calculator

    NSDL National Science Digital Library

    Stark, Philip B.

    This tool lets you calculate the probability that a random variable X is in a specified range, for a variety of probability distributions for X: the normal distribution, the binomial distribution with parameters n and p, the chi-square distribution, the exponential distribution, the geometric distribution, the hypergeometric distribution, the negative binomial distribution, the Poisson distribution, and Student's t-distribution. The first choice box lets you select a probability distribution. Depending on the distribution you select, text areas will appear for you to enter the values of the parameters of the distribution. Parameters that are probabilities (e.g., the chance of success in each trial for a binomial distribution) can be entered either as decimal numbers between 0 and 1, or as percentages. If you enter a probability as a percentage, be sure to include the percent sign (%) after the number.

  17. Development of a multicell methodology to account for heterogeneous core effects in the core-analysis diffusion code

    SciTech Connect

    Shen, W. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ont. L5K 1B2 (Canada)

    2006-07-01

    In CANDU R reactor calculations, the lattice-cell cross sections are calculated with WIMS-AECL, and the three-dimensional core neutron-flux and power distributions are calculated with RFSP-IST. The lattice-cell cross sections employed in RFSP-IST and in many other commercial core-analysis diffusion codes are usually based on the use of single-lattice-cell calculations, without considering the effects of the environment. This approximation is not sufficiently accurate for heterogeneous core configurations in the ACR-1000{sup TM}. A multicell correction method is therefore developed in RFSP-IST to account for heterogeneous core effects in the design and analysis of ACR-1000. The calculation results show that the multicell methodology developed in RFSP-IST is effective, generic, and it works well for ACR core analysis. (authors)

  18. 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.

  19. 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.

  20. Core and System Design of Reduced-Moderation Water Reactor with Passive Safety Features

    SciTech Connect

    Iwamura, Takamichi; Okubo, Tsutomu; Yonomoto, Taisuke [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 (Japan); Takeda, Renzo; Moriya, Kumiaki [Hitachi, Ltd. (Japan); Kanno, Minoru [The Japan Atomic Power Company (Japan)

    2002-07-01

    In order to ensure the sustainable energy supply in Japan, research and developments of reduced-moderation water reactor (RMWR) have been performed. The RMWR can attain the favorable characteristics such as high burn-up, long operation cycle, multiple recycling of plutonium and effective utilization of uranium resources, based on the matured LWR technologies. MOX fuel assemblies in the tight-lattice fuel rod arrangement are used to reduce the moderation of neutron, and hence, to increase the conversion ratio. The conceptual design has been accomplished for the small 330 MWe RMWR core with the discharge burn-up of 60 GWd/t and the operation cycle of 24 months, under the natural circulation cooling of the core. A breeding ratio of 1.01 and the negative void reactivity coefficient are simultaneously realized in the design. In the plant system design, the passive safety features are intended to be utilized mainly to improve the economy. At present, a hybrid one under the combination of the passive and the active components, and a fully passive one are proposed. The former has been evaluated to reduce the cost for the reactor components. (authors)

  1. hp calculators HP 50g Date calculations

    E-print Network

    Vetter, Frederick J.

    hp calculators HP 50g Date calculations The TIME menu Adding days to a date Days between dates Practice solving date problems #12;hp calculators HP 50g Date calculations hp calculators - 2 - HP 50g Date on the first level of the stack, prior to the execution of the DATE+ function. #12;hp calculators HP 50g Date

  2. Measurement of fission gas release, internal pressure and cladding creep rate in the fuel pins of PHWR bundle of normal discharge burnup

    Microsoft Academic Search

    U. K. Viswanathan; D. N. Sah; B. N. Rath; S. Anantharaman

    2009-01-01

    Fuel pins of a Pressurised Heavy Water Reactor (PHWR) fuel bundle discharged from Narora Atomic Power Station unit #1 after attaining a fuel burnup of 7528MWd\\/tU have been subjected to two types of studies, namely (i) puncture test to estimate extent of fission gas release and internal pressure in the fuel pin and (ii) localized heating of the irradiated fuel

  3. National Ice Core Laboratory

    NSDL National Science Digital Library

    USGS

    This facility stores, curates and studies ice cores recovered from glaciers from around the world. The site provides a photo gallery and description about each step of the process of drilling, transporting and analyzing the core. There is also a database of basic information about each core held at the laboratory and links to global change research information.

  4. Cores of combined games

    Microsoft Academic Search

    Francis Bloch; Geoffroy De Clippel

    2008-01-01

    This paper studies the core of combined games, obtained by summing two coalitional games. It is shown that the set of balanced transferable utility games can be partitioned into equivalence classes of component games whose core is equal to the core of the combined game. On the other hand, for non balanced games, the binary relation associating two component games

  5. Cores of Combined Games

    Microsoft Academic Search

    Francis Bloch; Geoffroy de Clippel

    2009-01-01

    This paper studies the core of combined games, obtained by summing two coalitional games. It is shown that the set of balanced transferable utility games can be partitioned into equivalence classes of component games to determine whether the core of the combined game coincides with the sum of the cores of its components. On the other hand, for non-balanced games,

  6. Evolution of the core physics concept for the Canadian supercritical water reactor

    SciTech Connect

    Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

    2013-07-01

    The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.

  7. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    SciTech Connect

    Merk, B. [Helmholtz-Zentrum Dresden-Rossendorf, Institut fuer Sicherheitsforschung, Postfach 51 01 19, 01314 Dresden (Germany); Weiss, F. P. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH Forschungszentrum, Boltzmannstr. 14, 85748 Garching (Germany)

    2012-07-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  8. Banded transformer cores

    NASA Technical Reports Server (NTRS)

    Mclyman, C. W. T. (inventor)

    1974-01-01

    A banded transformer core formed by positioning a pair of mated, similar core halves on a supporting pedestal. The core halves are encircled with a strap, selectively applying tension whereby a compressive force is applied to the core edge for reducing the innate air gap. A dc magnetic field is employed in supporting the core halves during initial phases of the banding operation, while an ac magnetic field subsequently is employed for detecting dimension changes occurring in the air gaps as tension is applied to the strap.

  9. Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

    SciTech Connect

    Eliseev, V.A.; Malisheva, I.V.; Matveev, V.I.; Egorov, A.V.; Maslov, P.A. [SSC RF - IPPE, Obninsk, Kaluga region (Russian Federation)

    2013-07-01

    With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.

  10. Compositional Model for the Earth's Core

    NASA Astrophysics Data System (ADS)

    McDonough, W. F.

    2003-12-01

    The remote setting of the Earth's core tests our ability to assess its physical and chemical characteristics. Extending out to half an Earth radii, the metallic core constitutes a sixth of the planet's volume and a third of its mass (see Table 1 for physical properties of the Earth's core). The boundary between the silicate mantle and the core (CMB) is remarkable in that it is a zone of greatest contrast in Earth properties. The density increase across this boundary represents a greater contrast than across the crust-ocean surface. The Earth's gravitational acceleration reaches a maximum (10.7 m s-2) at the CMB and this boundary is also the site of the greatest temperature gradient in the Earth. (The temperature at the base of the mantle (˜2,900 °C) is not well established, and that at the top of the inner core is even less securely known (˜3,500-4,500 °C).) The pressure range throughout the core (i.e., 136 GPa to >360 GPa) makes recreating environmental conditions in most experimental labs impossible, excepting a few diamond anvil facilities or those with high-powered, shock-melting guns (see Chapter 2.14). Thus, our understanding of the core is based on very few pieces of direct evidence and many fragments of indirect observations. Direct evidence comes from seismology, geodesy, geo- and paleomagnetism, and, relatively recently isotope geochemistry (see Section 2.15.6). Indirect evidence comes from geochemistry, cosmochemistry, and meteoritics; further constraints on the core system are gained from studies in experimental petrology, mineral physics, ab initio calculations, and evaluations of the Earth's energy budget (e.g., geodynamo calculations, core crystallization, heat flow across the core-mantle boundary). Figure 1 provides a synopsis of research on the Earth's core, and the relative relationship between disciplines. Feedback loops between all of these disciplines refine other's understanding of the Earth's core. Table 1. Physical properties of the Earth's core UnitsRefs. Mass Earth5.9736E+24kg1 Inner core9.675E+22kg1 Outer core1.835E+24kg1 Core1.932E+24kg1 Mantle4.043E+24kg1 Inner core to core (%)5.0% Core to Earth (%)32.3% Depth Core-mantle boundary3,483±5km2 Inner-outer core boundary1,220±10km2 Mean radius of the Earth6,371.01±0.02km1 Volume relative to planet Inner core7.606E+09(0.7%)km3 Inner core relative to the bulk core4.3% Outer core1.694E+11(15.6%)km3 Bulk core1.770E+11(16.3%)km3 Silicate earth9.138E+11(84%)km3 Earth1.083E+12km3 Moment of inertia constants Earth mean moment of inertia (I)0.3299765Ma21 Earth mean moment of inertia (I)0.3307144MR021 Mantle: Im/Ma20.29215Ma21 Fluid core: If/Ma20.03757Ma21 Inner core: Iic/Ma22.35E-4Ma21 Core: If+ic/Mf+icaf20.392Ma21 1 - Yoder (1995), 2 - Masters and Shearer (1995). M is the Earth's mass, a is the Earth's equatorial radius, R0 is the radius for an oblate spheroidal Earth, Im is the moment of inertia for the mantle, If is the moment of inertia for the outer (fluid) core, Iic is the moment of inertia for the inner core, and If+ic/Mf+icaf2 is the mean moment of inertia for the core. (11K)Figure 1. The relative relationship between disciplines involved in research on the Earth's core and the nature of data and information that come from these various investigations. Studies listed in the upper row yield direct evidence on properties of the core. Those in the middle row yield indirect evidence on the composition of the Earth's core, whereas findings from disciplines listed on the bottom row provide descriptions of the state conditions for the core and its formation.

  11. Updraft and Downdraft Cores in TOGA COARE: Why So Many Buoyant Downdraft Cores?.

    NASA Astrophysics Data System (ADS)

    Igau, Richard C.; Lemone, Margaret A.; Wei, Dingying

    1999-07-01

    An examination of the properties of updraft and downdraft cores using Electra data from TOGA COARE shows that they have diameters and vertical velocities similar to cores observed over other parts of the tropical and subtropical oceans. As in previous studies, a core is defined as having vertical velocity of the same sign and greater than an absolute value of 1 m s1 for at least 500 m. A requirement that the core contain either cloud or precipitation throughout is added, but this should not affect the results significantly.Since the Electra was equipped with the Ophir III radiometric temperature sensor, it was also possible to make estimates of core buoyancies. As in TAMEX and EMEX, where core temperatures were estimated using the modified side-looking Barnes radiometer on the NOAA P3s, a significant fraction of both updraft and downdraft cores had apparent virtual temperatures greater than their environments. In fact, the average virtual temperature deviation from the environment for downdraft cores was +0.4 K.Sixteen of the strongest downdraft cores were examined, all of which had positive virtual-temperature deviations, to find the source of this surprising result. It is concluded that the downdraft cores are artificially warm because 100% relative humidity was assumed in calculating virtual temperature. However, reducing core mixing ratios to more physically realistic values does not eliminate warm virtual potential temperature downdraft cores, nor does water loading make all cores negatively buoyant. Thus positively buoyant convective downdrafts do exist, though probably in smaller numbers than previously suggested.

  12. HYDRATE CORE DRILLING TESTS

    SciTech Connect

    John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

    2002-11-01

    The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate formation comprised of coarse, large-grain sand in ice. Results with this core showed that the viscosity of the drilling fluid must also be carefully controlled. When coarse sand was being cored, the core barrel became stuck because the drilling fluid was not viscous enough to completely remove the large grains of sand. These tests were very valuable to the project by showing the difficulties in coring permafrost or hydrates in a laboratory environment (as opposed to a field environment where drilling costs are much higher and the potential loss of equipment greater). Among the conclusions reached from these simulated hydrate coring tests are the following: Frozen hydrate core samples can be recovered successfully; A spring-finger core catcher works best for catching hydrate cores; Drilling fluid can erode the core and reduces its diameter, making it more difficult to capture the core; Mud must be designed with proper viscosity to lift larger cuttings; and The bottom 6 inches of core may need to be drilled dry to capture the core successfully.

  13. Neutronics calculations on the impact of burnable poisons to safety and non-proliferation aspects of inert matrix fuel

    NASA Astrophysics Data System (ADS)

    Pistner, C.; Liebert, W.; Fujara, F.

    2006-06-01

    Inert matrix fuels (IMF) with plutonium may play a significant role to dispose of stockpiles of separated plutonium from military or civilian origin. For reasons of reactivity control of such fuels, burnable poisons (BP) will have to be used. The impact of different possible BP candidates (B, Eu, Er and Gd) on the achievable burnup as well as on safety and non-proliferation aspects of IMF are analyzed. To this end, cell burnup calculations have been performed and burnup dependent reactivity coefficients (boron worth, fuel temperature and moderator void coefficient) were calculated. All BP candidates were analyzed for one initial BP concentration and a range of different initial plutonium-concentrations (0.4-1.0 g cm -3) for reactor-grade plutonium isotopic composition as well as for weapon-grade plutonium. For the two most promising BP candidates (Er and Gd), a range of different BP concentrations was investigated to study the impact of BP concentration on fuel burnup. A set of reference fuels was identified to compare the performance of uranium-fuels, MOX and IMF with respect to (1) the fraction of initial plutonium being burned, (2) the remaining absolute plutonium concentration in the spent fuel and (3) the shift in the isotopic composition of the remaining plutonium leading to differences in the heat and neutron rate produced. In the case of IMF, the remaining Pu in spent fuel is unattractive for a would be proliferator. This underlines the attractiveness of an IMF approach for disposal of Pu from a non-proliferation perspective.

  14. PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

    NASA Astrophysics Data System (ADS)

    Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

    2014-06-01

    The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

  15. The compactness of presupernova stellar cores

    SciTech Connect

    Sukhbold, Tuguldur; Woosley, S. E., E-mail: sukhbold@ucolick.org [Department of Astronomy and Astrophysics, University of California, Santa Cruz, CA 95064 (United States)

    2014-03-01

    The success or failure of the neutrino-transport mechanism for producing a supernova in an evolved massive star is known to be sensitive not only to the mass of the iron core that collapses, but also to the density gradient in the silicon and oxygen shells surrounding that core. Here we study the systematics of a presupernova core's 'compactness' as a function of the mass of the star and the physics used in its calculation. Fine-meshed surveys of presupernova evolution are calculated for stars from 15 to 65 M {sub ?}. The metallicity and the efficiency of semiconvection and overshoot mixing are both varied and bare carbon-oxygen cores are explored as well as full hydrogenic stars. Two different codes, KEPLER and MESA, are used for the study. A complex interplay of carbon and oxygen burning, especially in shells, can cause rapid variations in the compactness for stars of very nearly the same mass. On larger scales, the distribution of compactness with main sequence mass is found to be robustly non-monotonic, implying islands of 'explodabilty,' particularly around 8-20 M {sub ?} and 25-30 M {sub ?}. The carbon-oxygen (CO) core mass of a presupernova star is a better, (though still ambiguous) discriminant of its core structure than the main sequence mass.

  16. HD depletion in starless cores

    NASA Astrophysics Data System (ADS)

    Sipilä, O.; Caselli, P.; Harju, J.

    2013-06-01

    Aims: We aim to investigate the abundances of light deuterium-bearing species such as HD, H2D+, and D2H+ in a gas-grain chemical model that includes an extensive description of deuterium and spin-state chemistry, in physical conditions appropriate to the very centers of starless cores. Methods: We combined a gas-grain chemical model with radiative transfer calculations to simulate density and temperature structure in starless cores. The chemical model includes new reaction sets for both gas phase and grain surface chemistry, including deuterated forms of species with up to four atoms and the spin states of the light species H2, H2+, and H2+ and their deuterated forms. Results: We find that in the dense and cold environments attributed to the centers of starless cores, HD eventually depletes from the gas phase because deuterium is efficiently incorporated into grain-surface HDO, resulting in inefficient HD production on grains for advanced core ages. HD depletion has consequences not only on the abundances of, e.g., H2D+ and D2H+, whose production depends on the abundance of HD, but also on the spin state abundance ratios of the various light species, when compared with the complete depletion model where heavy elements do not influence the chemistry. Conclusions: While the eventual HD depletion leads to the disappearance of light deuterium-bearing species from the gas phase on a relatively short timescale at high density, we find that at late stages of core evolution, the abundances of H2D+ and D2H+ increase toward the core edge, and the distributions become extended. The HD depletion timescale increases if less oxygen is initially present in the gas phase, owing to chemical interaction between the gas and the dust preceding the starless core phase. Our results are greatly affected if H2 is allowed to tunnel on grain surfaces, and therefore more experimental data is needed not only on tunneling but also on the O + H2 surface reaction in particular.

  17. A Fission Gas Release Model for High-Burnup LWR ThOâ-UOâ Fuel

    Microsoft Academic Search

    Yun Long; Yi Yuan; Mujid S. Kazimi; Ronald G. Ballinger; Edward E. Pilat

    2002-01-01

    Fission gas release in thoria-urania fuel has been investigated by creating a specially modified FRAPCON-3 code. Because of the reduced buildup of ²³⁹Pu and a flatter distribution of ²³³U, the new model THUPS (Thoria-Urania Power Shape) was developed to calculate the radial power distribution, including the effects of both plutonium and ²³³U. Additionally, a new porosity model for the rim

  18. 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.

  19. The influence of the tape-core layer number of fluxgate sensor to the demagnetization factor

    Microsoft Academic Search

    Yulkifli; M. Djamal; Khairurrijal; D. Kurniadi; P. Ripka

    2009-01-01

    This paper explains the influence of the tape-core layer number to the demagnetization factor of a fluxgate sensor. The demagnetization factor was calculated based on the physical dimension, the self-inductance of coil without inserting the core (Lno_core) and by inserting the core (Lcore) of the sensor. The calculated demagnetization factor to pick-up coil configurations of 2 ?? 80 are proportional,

  20. Core sample extractor

    NASA Technical Reports Server (NTRS)

    Akins, James; Cobb, Billy; Hart, Steve; Leaptrotte, Jeff; Milhollin, James; Pernik, Mark

    1989-01-01

    The problem of retrieving and storing core samples from a hole drilled on the lunar surface is addressed. The total depth of the hole in question is 50 meters with a maximum diameter of 100 millimeters. The core sample itself has a diameter of 60 millimeters and will be two meters in length. It is therefore necessary to retrieve and store 25 core samples per hole. The design utilizes a control system that will stop the mechanism at a certain depth, a cam-linkage system that will fracture the core, and a storage system that will save and catalogue the cores to be extracted. The Rod Changer and Storage Design Group will provide the necessary tooling to get into the hole as well as to the core. The mechanical design for the cam-linkage system as well as the conceptual design of the storage device are described.

  1. Loss Evaluation in Converter Transformer Cores

    Microsoft Academic Search

    Yo Sakaki; Toshiro Sato; Tatsushi Ito

    1983-01-01

    A new iron loss evaluation parameter ¿ is presented to predict large signal iron losses in metallic and ferrite cores on the basis of the results of small signal measurements. Calculated results using ¿ is remarkablly in coincidence with the observed iron losses in the frequency range from several tens kHz to 1MHz. An experimental study on the relation between

  2. Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

    SciTech Connect

    G. S. Chang; M. A. Lillo; R. G. Ambrosek

    2008-06-01

    The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.

  3. A NEW METHOD TO QUANTIFY CORE TEMPERATURE INSTABILITY IN RODENTS.

    EPA Science Inventory

    Methods to quantify instability of autonomic systems such as temperature regulation should be important in toxicant and drug safety studies. Stability of core temperature (Tc) in laboratory rodents is susceptible to a variety of stimuli. Calculating the temperature differential o...

  4. Modelling of a novel hollow-core photonic crystal fibre

    Microsoft Academic Search

    T. D. Hedley; D. M. Bird; F. Benabid; J. C. Knight; P.St.J. Russell

    2003-01-01

    A hollow-core photonic crystal fibre that guides light over a broad range of frequencies has been developed. Calculations indicate that guidance is due to a low density of states of the cladding structure near the air line.

  5. Birefringent optical fiber with a photonic crystal core

    Microsoft Academic Search

    J. J. Hu; M. Yan; P. Shum

    2006-01-01

    Photonic crystals with reduced symmetry are introduced into a fiber's core to induce birefringence between the two nearly-orthogonal fundamental modes. We provide our preliminary findings on the birefringence property of such fibers through numerical calculations.

  6. Core excitations across the neutron shell gap in 207Tl

    NASA Astrophysics Data System (ADS)

    Wilson, E.; Podolyák, Zs.; Grawe, H.; Brown, B. A.; Chiara, C. J.; Zhu, S.; Fornal, B.; Janssens, R. V. F.; Shand, C. M.; Bowry, M.; Bunce, M.; Carpenter, M. P.; Cieplicka-Ory?czak, N.; Deo, A. Y.; Dracoulis, G. D.; Hoffman, C. R.; Kempley, R. S.; Kondev, F. G.; Lane, G. J.; Lauritsen, T.; Lotay, G.; Reed, M. W.; Regan, P. H.; Rodríguez Triguero, C.; Seweryniak, D.; Szpak, B.; Walker, P. M.

    2015-07-01

    The single closed-neutron-shell, one proton-hole nucleus 207Tl was populated in deep-inelastic collisions of a 208Pb beam with a 208Pb target. The yrast and near-yrast level scheme has been established up to high excitation energy, comprising an octupole phonon state and a large number of core excited states. Based on shell-model calculations, all observed single core excitations were established to arise from the breaking of the N = 126 neutron core. While the shell-model calculations correctly predict the ordering of these states, their energies are compressed at high spins. It is concluded that this compression is an intrinsic feature of shell-model calculations using two-body matrix elements developed for the description of two-body states, and that multiple core excitations need to be considered in order to accurately calculate the energy spacings of the predominantly three-quasiparticle states.

  7. Core Disruptive Accident Analysis using ASTERIA-FBR

    NASA Astrophysics Data System (ADS)

    Ishizu, Tomoko; Endo, Hiroshi; Yamamoto, Toshihisa; Tatewaki, Isao

    2014-06-01

    JNES is developing a core disruptive accident analysis code, ASTERIA-FBR, which tightly couples the thermal-hydraulics and the neutronics to simulate the core behavior during core disruptive accidents of fast breeder reactors (FBRs). ASTERIA-FBR consists of the three-dimensional thermal-hydraulics calculation module: CONCORD, the fuel pin behavior calculation module: FEMAXI-FBR, and the space-time neutronics module: Dynamic-GMVP or PARTISN/RKIN. This paper describes a comparison between characteristics of GMVP and PARTISN and summarizes the challenging issues on applying Dynamic-GMVP to the calculation against unprotected loss-of-flow (ULOF) event which is a typical initiator of core disruptive accident of FBR. The statistical error included in the calculation results may affect the super-prompt criticality during ULOF event and thus the amount of released energy.

  8. Magnetoconvection and Thermal Coupling of the Earth's Core and Mantle

    Microsoft Academic Search

    Peter Olson; Gary A. Glatzmaier

    1996-01-01

    Numerical calculations of finite amplitude magnetoconvection in a rotating spherical shell are used to investigate the thermal coupling of the Earth's core to the mantle. From the observed distribution of lower mantle seismic velocity heterogeneity we construct a pattern of heat flow on the core-mantle boundary consisting of a spherical average q0 plus heterogeneity with amplitude Delta q. For Delta

  9. An innovative optical and chemical drill core scanner

    NASA Astrophysics Data System (ADS)

    Sjöqvist, A. S. L.; Arthursson, M.; Lundström, A.; Calderón Estrada, E.; Inerfeldt, A.; Lorenz, H.

    2015-05-01

    We describe a new innovative drill core scanner that semi-automatedly analyses drill cores directly in drill core trays with X-ray fluorescence spectrometry, without the need for much sample preparation or operator intervention. The instrument is fed with entire core trays, which are photographed at high resolution and scanned by a 3-D profiling laser. Algorithms recognise the geometry of the core tray, number of slots, location of the drill cores, calculate the optimal scanning path, and execute a continuous XRF analysis of 2 cm width along the core. The instrument is equipped with critical analytical components that allow an effective QA/QC routine to be implemented. It is a mobile instrument that can be manoeuvred by a single person with a manual pallet jack.

  10. Iron diffusion from first principles calculations

    NASA Astrophysics Data System (ADS)

    Wann, E.; Ammann, M. W.; Vocadlo, L.; Wood, I. G.; Lord, O. T.; Brodholt, J. P.; Dobson, D. P.

    2013-12-01

    The cores of Earth and other terrestrial planets are made up largely of iron1 and it is therefore very important to understand iron's physical properties. Chemical diffusion is one such property and is central to many processes, such as crystal growth, and viscosity. Debate still surrounds the explanation for the seismologically observed anisotropy of the inner core2, and hypotheses include convection3, anisotropic growth4 and dendritic growth5, all of which depend on diffusion. In addition to this, the main deformation mechanism at the inner-outer core boundary is believed to be diffusion creep6. It is clear, therefore, that to gain a comprehensive understanding of the core, a thorough understanding of diffusion is necessary. The extremely high pressures and temperatures of the Earth's core make experiments at these conditions a challenge. Low-temperature and low-pressure experimental data must be extrapolated across a very wide gap to reach the relevant conditions, resulting in very poorly constrained values for diffusivity and viscosity. In addition to these dangers of extrapolation, preliminary results show that magnetisation plays a major role in the activation energies for diffusion at low pressures therefore creating a break down in homologous scaling to high pressures. First principles calculations provide a means of investigating diffusivity at core conditions, have already been shown to be in very good agreement with experiments7, and will certainly provide a better estimate for diffusivity than extrapolation. Here, we present first principles simulations of self-diffusion in solid iron for the FCC, BCC and HCP structures at core conditions in addition to low-temperature and low-pressure calculations relevant to experimental data. 1. Birch, F. Density and composition of mantle and core. Journal of Geophysical Research 69, 4377-4388 (1964). 2. Irving, J. C. E. & Deuss, A. Hemispherical structure in inner core velocity anisotropy. Journal of Geophysical Research 116, B04307 (2011). 3. Buffett, B. A. Onset and orientation of convection in the inner core. Geophysical Journal International 179, 711-719 (2009). 4. Bergman, M. Measurements of electric anisotropy due to solidification texturing and the implications for the Earth's inner core. Nature 389, 60-63 (1997). 5. Deguen, R. & Cardin, P. Thermochemical convection in Earth's inner core. Geophysical Journal International 187, 1101-1118 (2011). 6. Reaman, D. M., Daehn, G. S. & Panero, W. R. Predictive mechanism for anisotropy development in the Earth's inner core. Earth and Planetary Science Letters 312, 437-442 (2011). 7. Ammann, M. W., Brodholt, J. P., Wookey, J. & Dobson, D. P. First-principles constraints on diffusion in lower-mantle minerals and a weak D'' layer. Nature 465, 462-5 (2010).

  11. Azimuthal-spin-wave-mode-driven vortex-core reversals

    NASA Astrophysics Data System (ADS)

    Yoo, Myoung-Woo; Kim, Sang-Koog

    2015-01-01

    We studied, by micromagnetic numerical calculations, asymmetric vortex-core reversals driven by the m = -1 and m = +1 azimuthal spin-wave modes' excitations in soft magnetic circular nano-disks. We addressed the similarities and differences between the asymmetric core reversals in terms of the temporal evolutions of the correlated core-motion speed, locally concentrated perpendicular gyrofield, and magnetization dip near the original vortex core. The criterion for the core reversals was found to be the magnetization dip that must reach the out-of-plane magnetization component, mz = -p, with the initial polarization p, where p = +1 (-1) for the upward (downward) core magnetization. The core-motion speed and the associated perpendicular gyrofield, variable and controllable with static perpendicular field, Hz, applied perpendicularly to the disk plane, must reach their threshold values to meet the ultimate core-reversal criterion. Also, we determined the Hz strength and direction dependence of the core-switching time and threshold exciting field strength required for the core reversals, whose parameters are essential in the application aspect. This work offers deeper insights into the azimuthal spin-wave-driven core-reversal dynamics as well as an efficient means of controlling the azimuthal-modes-driven core reversals.

  12. Collapse and fragmentation of molecular cloud cores. I - Moderately centrally condensed cores

    NASA Technical Reports Server (NTRS)

    Boss, Alan P.

    1993-01-01

    3D calculations of the collapse of moderately centrally condensed molecular cloud cores with varied thermal and rotational energies are presented. The calculations are carried out using a newly developed and tested second-order accurate radiative hydrodynamics code. Because of the use of a second-order accurate numerical scheme and initial clouds that resemble both observed prolate molecular cloud cores and magnetically supported clouds at the initiation of the dynamic collapse phase, the new models provide a superior estimate of the likelihood of fragmentation as a mechanism for binary star formation.

  13. Powder Cores s Molypermalloy

    E-print Network

    Software q Current Transformer Design Software q Inductor Design Software q Mag Amp Design Software POWDER.mag-inc.com PRODUCT LITERATURE AND DESIGN SOFTWARE CD CONTAINS q All Product Literature q Common Mode FIlter Design the Proper Core for Saturating Transformers q TWC-S3 Inverter Transformer Core Design and Material Selection

  14. Core Knowledge Resource Guide.

    ERIC Educational Resources Information Center

    Core Knowledge Foundation, Charlottesville, VA.

    This resource guide is an annotated bibliography of books and other print sources relevant to the content specified in the Core Knowledge Sequences for grades 1 through 6, which are curricular guidelines published by the Core Knowledge Foundation. Entries are organized by subject area and grade level. Within each grade and subject area, entries…

  15. Reinforced sand cores

    Microsoft Academic Search

    Martin Zoldan

    2005-01-01

    Engine blocks and cylinder heads (castings) are made of aluminum or cast iron. Molten metal, poured into molds, forms the shape of engine blocks and cylinder heads. Molds create the outside of the casting and sand cores create cavities within the casting. ^ Typically, sand cores must maintain small aspect ratios to preserve structural integrity during the casting process. The

  16. Cores of convex games

    Microsoft Academic Search

    Lloyd S. Shapley

    1971-01-01

    The core of ann-person game is the set of feasible outcomes that cannot be improved upon by any coalition of players. A convex game is defined as one that is based on a convex set function. In this paper it is shown that the core of a convex game is not empty and that it has an especially regular structure.

  17. Secure Core Contact Information

    E-print Network

    Secure Core Contact Information C. E. Irvine irvine@nps.edu 831-656-2461 Department of Computer for the secure management of local and/or remote information in multiple contexts. The SecureCore project Science Graduate School of Operations and Information Sciences www.cisr.nps.edu Project Description

  18. Direct core materials.

    PubMed

    Wilson, Paul H; Fisher, Nigel L; Bartlett, David W

    2003-09-01

    There are many materials that can be used for direct-placement cores. Although the scientific evidence is incomplete, some materials are better suited to this task than others. This article provides an overview of direct-placement core materials and highlights what clinicians should consider when assessing a new product. PMID:14558201

  19. Chemical Models of Star-Forming Cores

    NASA Astrophysics Data System (ADS)

    Aikawa, Y.

    2013-10-01

    We review chemical models of low-mass star forming cores including our own work. Chemistry in molecular clouds are not in equilibrium. Molecular abundances in star forming cores change not only with physical conditions in cores but also with time. In prestellar cores, temperature stays almost constant ˜ 10 K, while the gas density increases as the core collapses. Three chemical phenomena are observed in this cold phase: molecular depletion, chemical fractionation, and deuterium enrichment. They are reproduced by chemical models combined with isothermal gravitational collapse. The collapse timescale of prestellar cores depends on the initial ratios of thermal, turbulent and magnetic pressure to gravitational energy. Since the chemical timescales, such as adsorption timescale of gas particle onto grains, are comparable to the collapse timescale, molecular abundances in cores should vary depending on the collapse timescale. Observations found that molecular abundances in some cores deviate from those in other cores, in spite of their similar central densities; it could originate in the pressure to gravity ratio in the cores. As the core contraction proceeds, compressional heating eventually overwhelms radiative cooling, and the core starts to warm up. Temperature of the infalling gas rises, as it approaches the central region. Grain-surface reactions of adsorbed molecules occur in this warm-up phase, as well as in prestellar phase. Hydrogenation is efficient at T ? 20 K, whereas radicals can migrate on grain surface and react with each other to form complex organic molecules (COMs) at T ? 30 K. Grain-surface species are sublimated to the gas phase and re-start gas-phase reactions; e.g. unsaturated carbon chains are formed from sublimated methane. Our model calculation predicts that COMs increases as the warm region extends outwards and the abundances of unsaturated carbon chains depend on the gas density in the CH4 sublimation zone. Recent detection of COMs in prestellar cores may indicate that a fraction of COMs formed in the vicinity of a protostar could be distributed to ambient clouds by outflows. COMs and carbon chains in protostellar phase inherit the high D/H ratio of their mother molecules, which originate mostly in cold prestellar phase.

  20. 34. DESPATCH CORE OVENS, GREY IRON FOUNDRY CORE ROOM, BAKES ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    34. DESPATCH CORE OVENS, GREY IRON FOUNDRY CORE ROOM, BAKES CORES THAT ARE NOT MADE ON HEATED OR COLD BOX CORE MACHINES, TO SET BINDING AGENTS MIXED WITH THE SAND CREATING CORES HARD ENOUGH TO WITHSTAND THE FLOW OF MOLTEN IRON INSIDE A MOLD. - Stockham Pipe & Fittings Company, Grey Iron Foundry, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  1. Qualification of the B and W Mark B fuel assembly for high burnup. First semi-annual progress report, July-December 1978

    Microsoft Academic Search

    T. A. Coleman; E. J. Coppola; P. L. Doss; V. O. Uotinen; H. H. Davis

    1979-01-01

    Five Babcock and Wilcox standard Mark B (15 x 15) fuel assemblies are being irradiated in Duke Power Company's Oconee Unit 1 reactor under a research and development program sponsored by the U.S. Department of Energy. Valuable experimental data on fuel performance characteristics at burnups of > 40,000 MWd\\/mtU will be obtained from these assemblies. This information, at a duty

  2. Determination of burnup in spent nuclear fuel by application of fiber optic high-resolution inductively coupled plasma atomic emission spectroscopy (FO-HR-ICP-AES)

    Microsoft Academic Search

    J. J Giglio; D. G Cummings; M. M Michlik; P. S Goodall; S. G Johnson

    1997-01-01

    The determination of burnup, an indicator of fuel cycle efficiency, has been accomplished by the determination of 139La 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 m

  3. EPMA and SEM of fuel samples from PWR rods with an average burn-up of around 100 MWd\\/kgHM

    Microsoft Academic Search

    R. Manzel; C. T Walker

    2002-01-01

    Samples from the high and low power regions of two fuel rods with average burn-ups of 89.5 and 97.8 MWd\\/kgHM were examined. Electron probe microanalysis (EPMA) was used to measure the radial distributions of Xe, Cs and Nd in the UO2 fuel matrix, and scanning electron microscopy (SEM) was used to study the change in the UO2 microstructure across the

  4. 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.

  5. Lunar Core and Tides

    NASA Technical Reports Server (NTRS)

    Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.

    2004-01-01

    Variations in rotation and orientation of the Moon are sensitive to solid-body tidal dissipation, dissipation due to relative motion at the fluid-core/solid-mantle boundary, and tidal Love number k2 [1,2]. There is weaker sensitivity to flattening of the core-mantle boundary (CMB) [2,3,4] and fluid core moment of inertia [1]. Accurate Lunar Laser Ranging (LLR) measurements of the distance from observatories on the Earth to four retroreflector arrays on the Moon are sensitive to lunar rotation and orientation variations and tidal displacements. Past solutions using the LLR data have given results for dissipation due to solid-body tides and fluid core [1] plus Love number [1-5]. Detection of CMB flattening, which in the past has been marginal but improving [3,4,5], now seems significant. Direct detection of the core moment has not yet been achieved.

  6. Collapse of a molecular cloud core to stellar densities: stellar-core and outflow formation in radiation magnetohydrodynamic simulations

    NASA Astrophysics Data System (ADS)

    Bate, Matthew R.; Tricco, Terrence S.; Price, Daniel J.

    2014-01-01

    We have performed smoothed particle radiation magnetohydrodynamic simulations of the collapse of rotating, magnetized molecular cloud cores to form protostars. The calculations follow the formation and evolution of the first hydrostatic core, the collapse to form a stellar core, the launching of outflows from both the first hydrostatic core and stellar core, and the breakout of the stellar outflow from the remnant of the first core. We investigate the roles of magnetic fields and thermal feedback on the outflow launching process, finding that both magnetic and thermal forces contribute to the launching of the stellar outflow. We also follow the stellar cores until they grow to masses of up to 20 Jupiter-masses, and determine their properties. We find that at this early stage, before fusion begins, the stellar cores have radii of ?3 R? with radial entropy profiles that increase outward (i.e. are convectively stable) and minimum entropies per baryon of s/kB ? 14 in their interiors. The structure of the stellar cores is found to be insensitive to variations in the initial magnetic field strength. With reasonably strong initial magnetic fields, accretion on to the stellar cores occurs through inspiralling magnetized pseudo-discs with negligible radiative losses, as opposed to first cores which effectively radiate away the energy liberated in the accretion shocks at their surfaces. We find that magnetic field strengths of >10 kG can be implanted in stellar cores at birth.

  7. Solvent extraction studies with low-burnup Fast Flux Test Facility fuel in the Solvent Extraction Test Facility

    SciTech Connect

    Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Chattin, F.R.; King, L.J.; Kitts, F.G.; Ross, R.G.; Stacy, R.G.

    1985-01-01

    A batch of irradiated Fast Flux Test Facility (FFTF) fuel was processed for the first time in the Solvent Extraction Test Facility (SETF) at the Oak Ridge National Laboratory (ORNL) during Campaign 7. The average burnup of the fuel was only 0.2 atom %, but the cooling time was short enough ({similar_to}2 years) so that {sup 95}Zr was detected in the feed. This short cooling permitted our first measurement of {sup 95}Zr decontamination factors (DFs) without having to use tracers. No operational problems were noted in the operation of the extraction-scrubbing contactor, and low uranium and plutonium losses (< 0.01%) were measured. Fission product DFs were improved noticeably by increasing the number of scrub stages from six to eight. Two flowsheet options for making pure uranium and plutonium products (total partitioning) were tested. Each flowsheet used hydroxylamine nitrate for reducing plutonium. Good products were obtained (uranium DFs of > 10{sup 3} and plutonium DFs of > 10{sup 4}), but each flowsheet was troubled with plutonium reoxidation. Adding hydrazine and lowering the plutonium concentration lessened the problem but did not eliminate it. About 370 g of plutonium was recovered from these tests, purified by anion exchange, converted to PuO{sub 2}, and transferred to the fuel refabrication program. 7 references.

  8. 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 [Institut fuer Nukleare Entsorgung, Forschungszentrum Karlsruhe, P.O.Box 3640, Karlsruhe, D- 76021 (Germany)

    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)

  9. 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.

  10. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    SciTech Connect

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

    2010-03-01

    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 Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm core characteristics, thermal-hydraulic properties, and radiation shielding. The high-temperature test operation at 950 ºC represented the fifth and final phase of the rise-to-power tests. The safety tests demonstrated inherent safety features of the HTTR such as slow temperature response during abnormal events due to the large heat capacity of the core and the negative reactivity feedback. The experimental benchmark performed and currently evaluated in this report pertains to the data available for the annular core criticals from the initial six isothermal, annular and fully-loaded, core critical measurements performed at the HTTR. Evaluation of the start-up core physics tests specific to the fully-loaded core is compiled elsewhere (HTTR-GCR-RESR-001).

  11. Compositional instability of Earth's solid inner core

    NASA Astrophysics Data System (ADS)

    Gubbins, D.; Alfè, D.; Davies, C. J.

    2013-03-01

    All models that invoke convection to explain the observed seismic variations in Earth's inner core require unstable inner core stratification. Previous work has assumed that chemical effects are stabilizing and focused on thermal convection, but recent calculations indicate that the thermal conductivity at core temperatures and pressures is so large that the inner core must cool entirely by conduction. We examine partitioning of oxygen, sulfur, and silicon in binary iron alloys and show that inner core growth results in a variable light element concentration with time: oxygen concentration decreases, sulfur concentration decreases initially and increases later, and silicon produces a negligible effect to within the model errors. The result is a net destabilizing concentration gradient. Convective stability is measured by a Rayleigh number, which exceeds the critical value for reasonable estimates of the viscosity and diffusivity. Our results suggest that inner core convection models, including the recently proposed translational mode, can be viable candidates for explaining seismic results if the driving force is compositional.

  12. Core-Polarization Contribution to the Nuclear Anapole Moment

    E-print Network

    N. Auerbach; B. A. Brown

    1999-03-11

    The importance of core contributions to the anapole moment in nuclei is examined. A model of the core-polarization correction is presented. The model is based on the coupling of the valence particles to the spin-dipole $J=1^{-}$ giant resonances of the core. A shell-model calculation of this correction is presented. The single-particle moments are calculated with Woods-Saxon and Skyrme Hartree Fock radial wave functions, and the general issues associated with nuclear configuration mixing are discussed.

  13. BWR AXIAL PROFILE

    SciTech Connect

    J. Huffer

    2004-09-28

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

  14. Plasma core at the center of a sonoluminescing bubble.

    PubMed

    Bemani, F; Sadighi-Bonabi, R

    2013-01-01

    Considering high temperature and pressure during single bubble sonoluminescence collapse, a hot plasma core is generated at the center of the bubble. In this paper a statistical mechanics approach is used to calculate the core pressure and temperature. A hydrochemical model alongside a plasma core is used to study the bubble dynamics in two host liquids of water and sulfuric acid 85 wt % containing Ar atoms. Calculation shows that the extreme pressure and temperature in the plasma core are mainly due to the interaction of the ionized Ar atoms and electrons, which is one step forward to sonofusion. The thermal bremsstrahlung mechanism of radiation is used to analyze the emitted optical energy per flash of the bubble core. PMID:23410423

  15. Plasma core at the center of a sonoluminescing bubble

    NASA Astrophysics Data System (ADS)

    Bemani, F.; Sadighi-Bonabi, R.

    2013-01-01

    Considering high temperature and pressure during single bubble sonoluminescence collapse, a hot plasma core is generated at the center of the bubble. In this paper a statistical mechanics approach is used to calculate the core pressure and temperature. A hydrochemical model alongside a plasma core is used to study the bubble dynamics in two host liquids of water and sulfuric acid 85 wt % containing Ar atoms. Calculation shows that the extreme pressure and temperature in the plasma core are mainly due to the interaction of the ionized Ar atoms and electrons, which is one step forward to sonofusion. The thermal bremsstrahlung mechanism of radiation is used to analyze the emitted optical energy per flash of the bubble core.

  16. Core assembly storage structure

    DOEpatents

    Jones, Jr., Charles E. (Northridge, CA); Brunings, Jay E. (Chatsworth, CA)

    1988-01-01

    A structure for the storage of core assemblies from a liquid metal-cooled nuclear reactor. The structure comprises an enclosed housing having a substantially flat horizontal top plate, a bottom plate and substantially vertical wall members extending therebetween. A plurality of thimble members extend downwardly through the top plate. Each thimble member is closed at its bottom end and has an open end adjacent said top plate. Each thimble member has a length and diameter greater than that of the core assembly to be stored therein. The housing is provided with an inlet duct for the admission of cooling air and an exhaust duct for the discharge of air therefrom, such that when hot core assemblies are placed in the thimbles, the heat generated will by convection cause air to flow from the inlet duct around the thimbles and out the exhaust duct maintaining the core assemblies at a safe temperature without the necessity of auxiliary powered cooling equipment.

  17. Biospecimen Core Resource

    Cancer.gov

    The Cancer Genome Atlas (TCGA) Biospecimen Core Resource centralized laboratory reviews and processes blood and tissue samples and their associated data using optimized standard operating procedures for the entire TCGA Research Network.

  18. INTEGRAL core programme

    NASA Technical Reports Server (NTRS)

    Gehrels, N.; Schoenfelder, V.; Ubertini, P.; Winkler, C.

    1997-01-01

    The International Gamma Ray Astrophysics Laboratory (INTEGRAL) mission is described with emphasis on the INTEGRAL core program. The progress made in the planning activities for the core program is reported on. The INTEGRAL mission has a nominal lifetime of two years with a five year extension option. The observing time will be divided between the core program (between 30 and 35 percent during the first two years) and general observations. The core program consists of three main elements: the deep survey of the Galactic plane in the central radian of the Galaxy; frequent scans of the Galactic plane in the search for transient sources, and pointed observations of several selected sources. The allocation of the observation time is detailed and the sensitivities of the observations are outlined.

  19. Core-Noise Research

    NASA Technical Reports Server (NTRS)

    Hultgren, Lennart S.

    2012-01-01

    This presentation is a technical summary of and outlook for NASA-internal and NASA-sponsored external research on core noise funded by the Fundamental Aeronautics Program Subsonic Fixed Wing (SFW) Project. Sections of the presentation cover: the SFW system-level noise metrics for the 2015 (N+1), 2020 (N+2), and 2025 (N+3) timeframes; SFW strategic thrusts and technical challenges; SFW advanced subsystems that are broadly applicable to N+3 vehicle concepts, with an indication where further noise research is needed; the components of core noise (compressor, combustor and turbine noise) and a rationale for NASA's current emphasis on the combustor-noise component; the increase in the relative importance of core noise due to turbofan design trends; the need to understand and mitigate core-noise sources for high-efficiency small gas generators; and the current research activities in the core-noise area, with additional details given about forthcoming updates to NASA's Aircraft Noise Prediction Program (ANOPP) core-noise prediction capabilities, two NRA efforts (Honeywell International, Phoenix, AZ and University of Illinois at Urbana-Champaign, respectively) to improve the understanding of core-noise sources and noise propagation through the engine core, and an effort to develop oxide/oxide ceramic-matrix-composite (CMC) liners for broadband noise attenuation suitable for turbofan-core application. Core noise must be addressed to ensure that the N+3 noise goals are met. Focused, but long-term, core-noise research is carried out to enable the advanced high-efficiency small gas-generator subsystem, common to several N+3 conceptual designs, needed to meet NASA's technical challenges. Intermediate updates to prediction tools are implemented as the understanding of the source structure and engine-internal propagation effects is improved. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Quiet-Aircraft Subproject aims to develop concepts and technologies to reduce perceived community noise attributable to aircraft with minimal impact on weight and performance. This reduction of aircraft noise is critical to enabling the anticipated large increase in future air traffic.

  20. Core bounce supernovae

    SciTech Connect

    Cooperstein, J.

    1987-01-01

    The gravitational collapse mechanism for Type II supernovae is considered, concentrating on the direct implosion - core bounce - hydrodynamic explosion picture. We examine the influence of the stiffness of the dense matter equation of state and discuss how the shock wave is formed. Its chances of success are determined by the equation of state, general relativistic effects, neutrino transport, and the size of presupernova iron core. 12 refs., 1 tab.

  1. Central core disease

    Microsoft Academic Search

    Heinz Jungbluth

    2007-01-01

    Central core disease (CCD) is an inherited neuromuscular disorder characterised by central cores on muscle biopsy and clinical\\u000a features of a congenital myopathy. Prevalence is unknown but the condition is probably more common than other congenital myopathies.\\u000a CCD typically presents in infancy with hypotonia and motor developmental delay and is characterized by predominantly proximal\\u000a weakness pronounced in the hip girdle;

  2. In-core detector activation rate for a PWR assembly

    SciTech Connect

    Todosow, M.; Eisenhart, L.D.

    1982-01-01

    The in-core detector system is the principal source of information for determining relative assembly powers, and maximum fuel rod powers in a reactor core. The detector signals are used in conjunction with pre-calculated factors, and appropriate normalizations, to obtain measured power values. Considerable reliance is placed on the accuracy of in-core detector inferred power distributions in reactor operations, and in the verification of calculational methods. The objective of this study was to compare results from standard design codes for the in-core detector activation rate (and the fission rate distribution in an assembly), to results obtained from a detailed calculation performed with a continuous energy Monte Carlo program with ENDF/B-V nuclear data.

  3. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  4. Modeling and design of a reload PWR core for a 48-month fuel cycle

    SciTech Connect

    McMahon, M.V.; Driscoll, M.J.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

    1997-05-01

    The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

  5. Emergency core cooling system

    DOEpatents

    Schenewerk, William E. (Sherman Oaks, CA); Glasgow, Lyle E. (Westlake Village, CA)

    1983-01-01

    A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

  6. MCNP LWR Core Generator

    SciTech Connect

    Fischer, Noah A. [Los Alamos National Laboratory

    2012-08-14

    The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.

  7. Realizing low loss air core photonic crystal fibers by exploiting an antiresonant core surround.

    PubMed

    Roberts, P; Williams, D; Mangan, B; Sabert, H; Couny, F; Wadsworth, W; Birks, T; Knight, J; Russell, P

    2005-10-01

    The modal properties of an air core photonic crystal fiber which incorporates an anti-resonant feature within the region that marks the transition between the air core and the crystal cladding are numerically calculated. The field intensity at the glass/air interfaces is shown to be reduced by a factor of approximately three compared to a fiber with more conventional core surround geometry. The reduced interface field intensity comes at the expense of an increased number of unwanted core interface modes within the band gap. When the interface field intensity is associated with modal propagation loss, the findings are in accord with recent measurements on fabricated fibers which incorporate a similar antiresonant feature. PMID:19498857

  8. Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations

    SciTech Connect

    Tippayakul, C.; Ivanov, K. [Pennsylvania State Univ., Univ. Park (United States); Misu, S. [AREVA NP GmbH, An AREVA and SIEMENS Company, Erlangen (Germany)

    2006-07-01

    This paper presents the improvements of MCOR, a Monte Carlo depletion code system for fuel assembly reference calculations. The improvements of MCOR were initiated by the cooperation between the Penn State Univ. and AREVA NP to enhance the original Penn State Univ. MCOR version in order to be used as a new Monte Carlo depletion analysis tool. Essentially, a new depletion module using KORIGEN is utilized to replace the existing ORIGEN-S depletion module in MCOR. Furthermore, the online burnup cross section generation by the Monte Carlo calculation is implemented in the improved version instead of using the burnup cross section library pre-generated by a transport code. Other code features have also been added to make the new MCOR version easier to use. This paper, in addition, presents the result comparisons of the original and the improved MCOR versions against CASMO-4 and OCTOPUS. It was observed in the comparisons that there were quite significant improvements of the results in terms of k{sub inf}, fission rate distributions and isotopic contents. (authors)

  9. Phase separation of penetrable core mixtures

    E-print Network

    R. Finken; J. -P. Hansen; A. A. Louis

    2001-10-18

    A two-component system of penetrable particles interacting via a gaussian core potential is considered, which may serve as a crude model for binary polymer solutions. The pair structure and thermodynamic properties are calculated within the random phase approximation (RPA) and the hypernetted chain (HNC) integral equation. The analytical RPA predictions are in semi-quantitative agreement with the numerical solutions of the HNC approximation, which itself is very accurate for gaussian core systems. A fluid-fluid phase separation is predicted to occur for a broad range of potential parameters. The pair structure exhibits a nontrivial clustering behaviour of the minority component. Similiar conclusions hold for the related model of parabolic core mixtures, which is frequently used in dissipative particle dynamics (DPD) simulations.

  10. DANDE: a linked code system for core neutronics\\/depletion analysis

    Microsoft Academic Search

    R. J. LaBauve; T. R. England; D. C. George; R. E. MacFarlane; W. B. Wilson

    1985-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of

  11. DANDE: a linked code system for core neutronics\\/depletion analysis

    Microsoft Academic Search

    R. J. LaBauve; T. R. England; D. C. George; R. E. MacFarlane; W. B. Wilson

    1986-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of

  12. FEM Simulation of Effect of Non-Uniform Air Gap on Apparent Permeability of Cut Cores

    Microsoft Academic Search

    Stan Zurek

    2012-01-01

    The paper describes results of two-dimensional finite element modeling of cut (gapped) magnetic cores. The influence of an irregular air gap on the apparent permeability of the core is calculated. Two magnetically soft materials are used: electrical steel M-27 and supermalloy (80% Ni). The results indicate that decreasing the minimum air gap between parts of the core can increase the

  13. Advanced Fabrication Technique and Thermal Performance Prediction of U-Mo/Zr-alloy Dispersion Fuel Pin for High Burnup PWR

    NASA Astrophysics Data System (ADS)

    Suwardi

    2010-06-01

    In recent years, a novel class of zirconium alloys having the melting temperature of 990-1160 K has been developed. Based on novel zirconium matrix alloys, high uranium content fuel pin with U-9Mo has been developed according to capillary impregnation technique. The pin shows it is thermal conductivity ranging from 18 to 22 w/m/K that is comparably higher than UO2 pellet pin. The paper presents the met-met fabrication and thermal performance analysis of the fuel in typical PWR. The fabrication consists of mixing UO2 powder or granules and a novel Zr-alloy powder having low melting point, filling the mixture in a cladding tube that one of its end has been plugged, heating the pin to above melting temperature of Zr-alloy for an hour, natural cooling and heat treating at 300 K for 12 hr. The thermal analysis takes into account the pore and temperature distribution and high burn up effect to pellet conductivity. The thermal diffusivity ratio of novel to conventional fuel has been used as correction factor for the novel fuel conductivity. The results show a significant lowering pellet temperature along the radius until 1000 K at the hottest position. The analysis underestimates since the gap conductivity has been treated as decreased by 2% fission gas released that is not real since the use of lower temperature, and also decreasing thermal conductivity by porosity formation will much lower. The analysis shows that the novel fuel has very good thermal properties which able to pass the barrier of 65 MWD/kg-U, the limit to day commercial fuel. The burn-up extension means fewer fresh fuel is needed to produce electricity, preserve natural uranium resource, easier fuel handling operational per energy produced

  14. 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.

    2014-02-10

    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 500?C 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.

  15. Engineering Technology Core (ET Core) Guide

    NSDL National Science Digital Library

    "The ET Core is designed to prepare students for the study of courses specific to any engineering technology major. The curriculum provides hands-on work with technology and workplace relevance as students complete their study of physics, communications, and mathematics (through introductory calculus)." In this 140-page PDF, visitors will find an introduction to the course, the competencies it covers, equipment needed, and detailed instructions for all sixteen modules. The modules cover all sorts of engineering technology including Electrical, Thermal, Mechanical, Fluids, Optics, and Materials. Each module also contains any students handouts necessary to teach it.

  16. Industrial Technology Core (IT Core) Guide

    NSDL National Science Digital Library

    This resource, created by the South Carolina Advanced Technological Education (SC ATE) National Resource Center, introduces students to core projects of industrial technology. The lesson involves five different activities, the topics include: an introduction to technology careers, basic hand tools, mechanical advantage, basic electricity and hydraulic systems. A suggested equipment list, instructors notes, and objectives are included to guide instructors in preparing these lessons plans. Each one of these topics includes a worksheet for students to actively participate in these lessons. This is a comprehensive set of lessons to help students better understand the different elements in industrial technology.

  17. All-University Core Curriculum All-University Core Curriculum

    E-print Network

    Collett Jr., Jeffrey L.

    /gtPathways/curri culum.html _______________ 2.3 Page 1 #12;All-University Core Curriculum B. Mathematics.1 The objectiveAll-University Core Curriculum All-University Core Curriculum Office of Vice Provost-UNIVERSITY CORE CURRICULUM (AUCC) All Colorado State University students share a learning experience in common

  18. All-University Core Curriculum All-University Core Curriculum

    E-print Network

    All-University Core Curriculum _______________ 2.3 Page 1 All-University Core Curriculum Office of Vice Provost for Undergraduate Affairs Administration Building, Room 108 core.colostate.edu ALL-UNIVERSITY CORE CURRICULUM (AUCC) All Colorado State University students share a learning experience in common

  19. Pressure Core Characterization

    NASA Astrophysics Data System (ADS)

    Santamarina, J. C.

    2014-12-01

    Natural gas hydrates form under high fluid pressure and low temperature, and are found in permafrost, deep lakes or ocean sediments. Hydrate dissociation by depressurization and/or heating is accompanied by a multifold hydrate volume expansion and host sediments with low permeability experience massive destructuration. Proper characterization requires coring, recovery, manipulation and testing under P-T conditions within the stability field. Pressure core technology allows for the reliable characterization of hydrate bearing sediments within the stability field in order to address scientific and engineering needs, including the measurement of parameters used in hydro-thermo-mechanical analyses, and the monitoring of hydrate dissociation under controlled pressure, temperature, effective stress and chemical conditions. Inherent sampling effects remain and need to be addressed in test protocols and data interpretation. Pressure core technology has been deployed to study hydrate bearing sediments at several locations around the world. In addition to pressure core testing, a comprehensive characterization program should include sediment analysis, testing of reconstituted specimens (with and without synthetic hydrate), and in situ testing. Pressure core characterization technology can be used to study other gas-charged formations such as deep sea sediments, coal bed methane and gas shales.

  20. Core Noise Reduction

    NASA Technical Reports Server (NTRS)

    Hultgren, Lennart S.

    2011-01-01

    This presentation is a technical summary of and outlook for NASA-internal and NASA-sponsored external research on core (combustor and turbine) noise funded by the Fundamental Aeronautics Program Subsonic Fixed Wing (SFW) Project. Sections of the presentation cover: the SFW system-level noise metrics for the 2015, 2020, and 2025 timeframes; turbofan design trends and their aeroacoustic implications; the emerging importance of core noise and its relevance to the SFW Reduce-Perceived-Noise Technical Challenge; and the current research activities in the core noise area. Recent work1 on the turbine-transmission loss of combustor noise is briefly described, two2,3 new NRA efforts in the core-noise area are outlined, and an effort to develop CMC-based acoustic liners for broadband noise reduction suitable for turbofan-core application is delineated. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The reduction of aircraft noise is critical to enabling the anticipated large increase in future air traffic. The Subsonic Fixed Wing Project's Reduce-Perceived-Noise Technical Challenge aims to develop concepts and technologies to dramatically reduce the perceived aircraft noise outside of airport boundaries.

  1. Thermal mass limit of neutron cores

    NASA Astrophysics Data System (ADS)

    Roupas, Zacharias

    2015-01-01

    Static thermal equilibrium of a quantum self-gravitating ideal gas in general relativity is studied at any temperature, taking into account the Tolman-Ehrenfest effect. Thermal contribution to the gravitational stability of static neutron cores is quantified. The curve of maximum mass with respect to temperature is reported. At low temperatures the Oppenheimer-Volkoff calculation is recovered, while at high temperatures the recently reported classical gas calculation is recovered. An ultimate upper mass limit M =2.43 M? of all maximum values is found to occur at Tolman temperature T =1.27 mc2 with radius R =15.2 km .

  2. CORE LOSS ANALYSIS OF THE RELUCTANCE SYNCHRONOUS MOTOR WITH BARRIER ROTORS

    Microsoft Academic Search

    Peter Hudak; Valeria Hrabovcova; Pavol Rafajdus; Jozef Mihok

    2004-01-01

    This paper presents a numerical method for the core loss calculation based on the FEM. Magnitudes of the magnetic flux density in individual segments of the machine are gained by means of FEM. On the basis of Fourier analysis of the flux density waveform the core loss is calculated for a given value of voltage or current. This method is

  3. Journals in the Core Collection

    Microsoft Academic Search

    Thomas E. Nisonger

    2007-01-01

    This overview of the core concept applied to journals defines the relevant terminology and cites specific examples of core lists. Ten approaches for determining core journals (subjective judgment, use, indexing coverage, overlapping library holdings, citation data, citation network\\/co-citation analysis, production of articles, Bradford's Law, faculty publication data, and multiple criteria methods) are reviewed and the practical applications of core journals

  4. Martian thermal history, core segregation, and tectonics

    NASA Technical Reports Server (NTRS)

    Davies, G. F.; Arvidson, R. E.

    1981-01-01

    A series of calculated thermal histories of Mars is presented, and their possible relation to surface tectonic history is discussed. The models include convective heat transport through an empirical approximation, and heating by radioactivity and core segregation. Initial temperature and the timing and duration of core segregation are treated as free parameters. The initial temperature is the main determinant of Martian thermal evolution: as it is varied from 20 to 100% of the present mean temperature, the maximum in surface heat flux moves from very recent to very early in Martian history. For the latter cases, the details of core segregation control the detailed timing of a peak in the thermal flux that exceeded 100 mW/sq m. It is suggested that the early disruption of cratered terrain crust in the northern hemisphere and subsequent volcanic resurfacing may have been related to core segregation. This would be consistent with a scenario in which an early period of core segregation generated a marked peak in the thermal flux that may have led to extensive partial melting and volcanism. This scenario would require Mars to have had an initial mean temperature comparable to the present value.

  5. Rotational Coupling of the Pinned Core Superfluid

    E-print Network

    Jahan-Miri, M

    2010-01-01

    The effects of pinning between fluxoids and vortices in the core of a neutron star, on the dynamics of the core neutron superfluid are considered. The pinning impedes, but does not absolutely block, any radial as well as {\\em azimuthal} motion of the neutron vortices with respect to the lattice of fluxoids. The time scale for the coupling of rotation of the core superfluid to the rest of the star is calculated, allowing for the effect of the finite frictional force on the neutron vortices due to their pinning with the fluxoids. This turns out to be the dominant mechanism for the coupling of the core of a neutron star to its crust, as compared to the role of electron scattering, for most cases of interest. Furthermore, different behaviors for the post-glitch response of the core superfluid are distinguished that might be tested against the relevant observational data. Also, a conceptually important case (and controversial too, in the earlier studies on the role of the crustal superfluid) is realized where a su...

  6. Rotational Coupling of the Pinned Core Superfluid

    NASA Astrophysics Data System (ADS)

    Jahan-Miri, M.

    2010-12-01

    The effects of pinning between fluxoids and vortices in the core of a neutron star on the dynamics of the core neutron superfluid are considered. The pinning impedes, but does not absolutely block, any radial as well as azimuthal motion of the neutron vortices with respect to the lattice of fluxoids. The timescale for the coupling of rotation of the core superfluid to the rest of the star is calculated, allowing for the effect of the finite frictional force on the neutron vortices due to their pinning with the fluxoids. This turns out to be the dominant mechanism for the coupling of the core of a neutron star to its crust, as compared to the role of electron scattering, for most cases of interest. Furthermore, different behaviors for the post-glitch response of the core superfluid are distinguished that might be tested against the relevant observational data. Also, a conceptually important case (and controversial too, in the earlier studies on the role of the crustal superfluid) is realized where a superfluid may remain decoupled in spite of a spinning-up of its vortices.

  7. USGS National Ice Core Laboratory

    NSDL National Science Digital Library

    This United States Geological Survey site highlights the work of the National Ice Core Laboratory (NICL). It discusses the NICL's role, why ice cores are important to study, how ice cores are drilled and studied, and core drilling locations. These cores are recovered and studied for a variety of scientific investigations, most of which focus on the reconstruction of Earth's climate history. The facility currently houses over 14,000 meters of ice cores from 34 drill sites in Greenland, Antarctica, and high mountain glaciers in the Western United States. There are links for more information and individual core information such as numbers, locations and sizes.

  8. Air ingression calculations for selected plant transients using MELCOR

    SciTech Connect

    Kmetyk, L.N.

    1994-01-01

    Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition to the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression.

  9. Autistic Savant Calendar Calculators.

    ERIC Educational Resources Information Center

    Patti, Paul J.

    This study identified 10 savants with developmental disabilities and an exceptional ability to calculate calendar dates. These "calendar calculators" were asked to demonstrate their abilities, and their strategies were analyzed. The study found that the ability to calculate dates into the past or future varied widely among these calculators. Three…

  10. Arts at the Core

    NSDL National Science Digital Library

    The Arts at the Core Initiative is part of The College Board's Advocacy & Policy Center, created "to help transform education in America." Part of the Center's work involves the Arts at the Core project, whose goal is "to empower education leaders, particularly in under-resourced districts, to implement rigorous arts programming in their schools." Under the Our Progress section, visitors learn about some of the resources created to achieve this goal. Moving on, the News & Events area contains links to recent success stories of bringing arts education programs to schools, along with updates from the field of research into this area. Visitors shouldn't miss the Publications area, which contains a brochure about flagship programs and a summary of key recommendations for school systems seeking to move arts to the core of their mission.

  11. Shear viscosity in neutron star cores

    E-print Network

    P. S. Shternin; D. G. Yakovlev

    2008-08-21

    We calculate the shear viscosity $\\eta = \\eta_{e\\mu}+\\eta_{n}$ in a neutron star core composed of nucleons, electrons and muons ($\\eta_{e\\mu}$ being the electron-muon viscosity, mediated by collisions of electrons and muons with charged particles, and $\\eta_{n}$ the neutron viscosity, mediated by neutron-neutron and neutron-proton collisions). Deriving $\\eta_{e\\mu}$, we take into account the Landau damping in collisions of electrons and muons with charged particles via the exchange of transverse plasmons. It lowers $\\eta_{e\\mu}$ and leads to the non-standard temperature behavior $\\eta_{e\\mu}\\propto T^{-5/3}$. The viscosity $\\eta_{n}$ is calculated taking into account that in-medium effects modify nucleon effective masses in dense matter. Both viscosities, $\\eta_{e\\mu}$ and $\\eta_{n}$, can be important, and both are calculated including the effects of proton superfluidity. They are presented in the form valid for any equation of state of nucleon dense matter. We analyze the density and temperature dependence of $\\eta$ for different equations of state in neutron star cores, and compare $\\eta$ with the bulk viscosity in the core and with the shear viscosity in the crust.

  12. Core Composition and the Magnetic Field of Mercury

    NASA Astrophysics Data System (ADS)

    Spohn, T.; Breuer, D.

    2005-05-01

    The density of Mercury suggests a core of approximately 1800 km radius and a mantle of approximately 600 km thickness. Convection in the mantle is often claimed to be capable of freezing the core over the lifetime of the solar system if the core is nearly pure iron. The thermal history calculations of Stevenson et al. (1983) and Schubert et al. (1988) suggest that about 5 weight-% sulphur are required to lower the core liquidus sufficiently to prevent complete freezing of the core and maintain a significant fluid outer core shell. Other candidates for a light alloying element require similarly large concentrations. The requirement of a significant concentration of volatile elements in the core is likely to be at variance with cosmochemical arguments for a mostly refractory, volatile poor composition of the planet. We have re-addressed the question of the freezing of Mercury's core using parameterized convection models based on the stagnant lid theory of planetary mantle convection. We have compared these results to earlier calculations (Conzelmann and Spohn, 1999) of Hermian mantle convection using a finite-amplitude convection code. We find consistently that the stagnant lid tends to thermally insulate the deep interior and we find mantle and core temperatures significantly larger than those calculated by Stevenson et al. (1983) and Schubert et al. (1988). As a consequence we find fluid outer core shells for reasonable mantle rheology parameters even for compositions with as little as 0.1 weight-% sulphur. Stevenson, D.J., T. Spohn, and G. Schubert. Icarus, 54, 466, 1983. Schubert, G. M.N. Ross, D.J. Stevenson, and T. Spohn, in Mercury, F. Vilas, C.R. Chapman and M.S. Matthews, eds., p.429, 1988. Conzelmann, V. and T. Spohn, Bull. Am. Astr. Soc., 31, 1102, 1999.

  13. Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

    Microsoft Academic Search

    Yasunori Bessho; Osamu Yokomizo; Yuichiro Yoshimoto; Ryutaro Yamashita; Masumi Ishikawa; Akio Toba

    1997-01-01

    Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory

  14. Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges

    SciTech Connect

    B. McLeod

    2002-02-28

    This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M&O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M&O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report summarizes the results of the 2002 Reference SNF Discharge Projection.

  15. Superconducting tin core fiber

    NASA Astrophysics Data System (ADS)

    Homa, Daniel; Liang, Yongxuan; Hill, Cary; Kaur, Gurbinder; Pickrell, Gary

    2015-01-01

    In this study, we demonstrated superconductivity in a fiber with a tin core and fused silica cladding. The fibers were fabricated via a modified melt-draw technique and maintained core diameters ranging from 50-300 microns and overall diameters of 125-800 microns. Superconductivity of this fiber design was validated via the traditional four-probe test method in a bath of liquid helium at temperatures on the order of 3.8 K. The synthesis route and fiber design are perquisites to ongoing research dedicated all-fiber optoelectronics and the relationships between superconductivity and the material structures, as well as corresponding fabrication techniques.

  16. Global Ice Core Research

    NSDL National Science Digital Library

    This informative site from the US Geological Survey (USGS) covers the latest ice-core research projects from around the world, including sites in Nepal, Norway, and Kyrghyzstan. Authored by researchers at the Global Ice core Research Office, the site contains an overview of the mid-latitude and polar glaciers, isotopic methods in glacial research, and applications to paleoclimatology. Links to maps, figures, and in some cases, full-text articles (HTML) about specific glaciers are available, and the site is peppered with color photos of glacial environments. Links to biographies of the scientists involved in the project, contacts, and other snow and ice sites are also listed.

  17. Development of Safeguards System Simulator Composed of Multi-Functional Cores

    NASA Astrophysics Data System (ADS)

    Suzuki, Mitsutoshi; Ihara, Hitoshi

    Due to the large plutonium (Pu) throughput and high burn-up fuel in an advanced reprocessing facility, we are faced with the inevitable increasing burden of nuclear material accountancy (NMA) to meet the International Atomic Energy Agency (IAEA) safeguards criteria. A large volume of sampling analysis and inspectors' activities result in a great cost for facility operation. Therefore, it is increasingly important to evaluate a cost-effective performance for the safeguards system. In order to design an advanced safeguards system, we have initiated the development of a safeguards system simulator. The simulator is composed of several interrelated cores and a separate core is planned to develop. The NMA core is a near-real-time accounting (NRTA) code that had been originally developed more than ten years ago and has been improved on an objective-driven pre- and post-processor. A multivariate and multi-scale core based on a principle component analysis with a wavelet technique has been developed to provide an algorithm of process monitoring. The time and frequency decomposition was verified to be an effective technique to detect an abnormal event. In addition, a multiple optimization core has been developed with a fuzzy-linear-programming technique to investigate the cost-effective performance of the conceptual safeguards system. It is shown that a combination of flow-meter and non-destructive assay can be applied to the system in a cost-effective manner. In the future, a virtual design core will be developed to support a walk-through and three dimensional visible plant model.

  18. Electromagnetic pump stator core

    DOEpatents

    Fanning, Alan W. (San Jose, CA); Olich, Eugene E. (Aptos, CA); Dahl, Leslie R. (Livermore, CA)

    1995-01-01

    A stator core for supporting an electrical coil includes a plurality of groups of circumferentially abutting flat laminations which collectively form a bore and perimeter. A plurality of wedges are interposed between the groups, with each wedge having an inner edge and a thicker outer edge. The wedge outer edges abut adjacent ones of the groups to provide a continuous path around the perimeter.

  19. Utah's New Mathematics Core

    ERIC Educational Resources Information Center

    Utah State Office of Education, 2011

    2011-01-01

    Utah has adopted more rigorous mathematics standards known as the Utah Mathematics Core Standards. They are the foundation of the mathematics curriculum for the State of Utah. The standards include the skills and understanding students need to succeed in college and careers. They include rigorous content and application of knowledge and reflect…

  20. Ultrasonic Drilling and Coring

    NASA Technical Reports Server (NTRS)

    Bar-Cohen, Yoseph

    1998-01-01

    A novel drilling and coring device, driven by a combination, of sonic and ultrasonic vibration, was developed. The device is applicable to soft and hard objects using low axial load and potentially operational under extreme conditions. The device has numerous potential planetary applications. Significant potential for commercialization in construction, demining, drilling and medical technologies.

  1. Soil Core Sample #2

    USGS Multimedia Gallery

    Soil core obtained from existing goose grazing lawn along the Smith River in the Teshekpuk Lake Special Area of the National Petroleum Reserve - Alaska.  Buried peat layer broken open.  Closer examination of the buried peat layer demonstrates that non-salt-tolerant vegetation from the past...

  2. Soil Core Sample #1

    USGS Multimedia Gallery

    Soil core obtained from existing goose grazing lawn along the Smith River in the Teshekpuk Lake Special Area of the National Petroleum Reserve - Alaska.  The buried layer of peat beneath goose grazing lawn demonstrates that vegetation change has occurred in this area....

  3. From Context to Core

    ERIC Educational Resources Information Center

    Campus Technology, 2008

    2008-01-01

    At Campus Technology 2008, Arizona State University Technology Officer Adrian Sannier mesmerized audiences with his mandate to become more efficient by doing only the "core" tech stuff--and getting someone else to slog through the context. This article presents an excerpt from Sannier's hour-long keynote address at Campus Technology '08. Sannier…

  4. Resolving Supercritical Orion Cores

    NASA Astrophysics Data System (ADS)

    Li, Di; Chapman, N.; Goldsmith, P.; Velusamy, T.

    2009-01-01

    The theoretical framework for high mass star formation (HMSF) is unclear. Observations reveal a seeming dichotomy between high- and low-mass star formation, with HMSF occurring only in Giant Molecular Clouds (GMC), mostly in clusters, and with higher star formation efficiencies than low-mass star formation. One crucial constraint to any theoretical model is the dynamical state of massive cores, in particular, whether a massive core is in supercritical collapse. Based on the mass-size relation of dust emission, we select likely unstable targets from a sample of massive cores (Li et al. 2007 ApJ 655, 351) in the nearest GMC, Orion. We have obtained N2H+ (1-0) maps using CARMA with resolution ( 2.5", 0.006 pc) significantly better than existing observations. We present observational and modeling results for ORI22. By revealing the dynamic structure down to Jeans scale, CARMA data confirms the dominance of gravity over turbulence in this cores. This work was performed by the Jet Propulsion Laboratory, California Institute of Technology, under contract with the National Aeronautics and Space Administration.

  5. Deep Sea Coring

    NSDL National Science Digital Library

    Woods Hole Oceanographic Institution; Ocean and Climate Change Institute

    This Ocean and Climate Change Institute module features a brief, but image-rich overview of ocean drilling and sediment analysis to determine paleoclimate (past climate). This site is the first of a 3-page module, the other two sites (Describing the Core; Sampling Techniques) are linked at the top of the article.

  6. Navagating the Common Core

    ERIC Educational Resources Information Center

    McShane, Michael Q.

    2014-01-01

    This article presents a debate over the Common Core State Standards Initiative as it has rocketed to the forefront of education policy discussions around the country. The author contends that there is value in having clear cross state standards that will clarify the new online and blended learning that the growing use of technology has provided…

  7. Renewing the Core Curriculum

    ERIC Educational Resources Information Center

    Lawson, Hal A.

    2007-01-01

    The core curriculum accompanied the development of the academic discipline with multiple names such as Kinesiology, Exercise and Sport Science, and Health and Human Performance. It provides commonalties for undergraduate majors. It is timely to renew this curriculum. Renewal involves strategic reappraisals. It may stimulate change or reaffirm the…

  8. Internal core tightener. [LMFBR

    Microsoft Academic Search

    G. V. Brynsvold; H. J. Jr. Snyder

    1976-01-01

    An internal core tightener is disclosed which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal

  9. Internal core tightener

    Microsoft Academic Search

    Glen V. Brynsvold; Snyder Jr. Harold J

    1976-01-01

    An internal core tightener which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal operation involved

  10. Core competence (knowledge) (skill)

    E-print Network

    Wu, Yih-Min

    Core competence 8 5~8 2 3 4 5 6 7 8 PPS003 Ver. 1.1 2011/03/07 #12; 2 (knowledge) (skill) (attitude) Set of skill, knowledge or attitude which should be learned or acquired by each, 2000) (knowledge) (skill) (attitude) Set of skill, knowledge or attitude which should be learned

  11. Diagnostics of core barrel vibrations by in-core and ex-core neutron noise

    Microsoft Academic Search

    V. Arzhanov; I. Pázsit

    2003-01-01

    Diagnostics of core-barrel vibrations has traditionally been made by use of ex-vessel neutron detector signals. We suggest that in addition to the ex-core noise, also the in-core noise, induced by core barrel vibrations, be also used. This would enhance the possibilities of diagnostics where the number of the ex-core detectors is not sufficient or their positions are disadvantageous for effective

  12. Numerical models of the Earth’s thermal history: Effects of inner-core solidification and core potassium

    NASA Astrophysics Data System (ADS)

    Butler, S. L.; Peltier, W. R.; Costin, S. O.

    2005-09-01

    Recently there has been renewed interest in the evolution of the inner core and in the possibility that radioactive potassium might be found in significant quantities in the core. The arguments for core potassium come from considerations of the age of the inner core and the energy required to sustain the geodynamo [Nimmo, F., Price, G.D., Brodholt, J., Gubbins, D., 2004. The influence of potassium on core and geodynamo evolution. Geophys. J. Int. 156, 363-376; Labrosse, S., Poirier, J.-P., Le Mouël, J.-L., 2001. The age of the inner core. Earth Planet Sci. Lett. 190, 111-123; Labrosse, S., 2003. Thermal and magnetic evolution of the Earth's core. Phys. Earth Planet Int. 140, 127-143; Buffett, B.A., 2003. The thermal state of Earth's core. Science 299, 1675-1677] and from new high pressure physics analyses [Lee, K., Jeanloz, R., 2003. High-pressure alloying of potassium and iron: radioactivity in the Earth's core? Geophys. Res. Lett. 30 (23); Murthy, V.M., van Westrenen, W., Fei, Y.W., 2003. Experimental evidence that potassium is a substantial radioactive heat source in planetary cores. Nature 423, 163-165; Gessmann, C.K., Wood, B.J., 2002. Potassium in the Earth's core? Earth Planet Sci. Lett. 200, 63-78]. The Earth's core is also located at the lower boundary of the convecting mantle and the presence of radioactive heat sources in the core will affect the flux of heat between these two regions and will, as a result, have a significant impact on the Earth's thermal history. In this paper, we present Earth thermal history simulations in which we calculate fluid flow in a spherical shell representing the mantle, coupled with a core of a given heat capacity with varying degrees of internal heating in the form of K40 and varying initial core temperatures. The mantle model includes the effects of the temperature dependence of viscosity, decaying radioactive heat sources, and mantle phase transitions. The core model includes the thermal effects of inner core solidification and we present models for which the final size of the inner core is the same that for the present-day Earth. We compare the results of simulations with and without the effects of inner core solidification and we compare the results of the numerical model with those of a parameterized model. Models with concentrations of potassium in the core of roughly 600 ppm best satisfy the present-day surface heat flow constraint; however, the core temperatures in these models are somewhat high. In addition, we find that models with lesser degrees of heating in the core can also satisfy the surface heat flow constraint provided that the mantle is in a particularly active state. Our models predict a relatively young inner core with the greatest age being 1756 Ma. We demonstrate that models with high core temperatures in the latter part of simulations result in high CMB heat flows which lead to predictions of young inner cores. For fixed initial core temperatures, this leads to a slight decrease in the predicted age of the inner core with increasing concentration of radioactive elements in the core.

  13. The Calculator Reference

    NSDL National Science Digital Library

    Furr, Rick.

    If you remember the days when calculators were as big as today's laptops, then you'll definitely feel nostalgic when you visit the Calculator Reference. Operated by an avid collector of vintage calculators, the site covers Texas Instruments and Hewlett-Packard brands, as well as the pioneering Curta. Several other models are included in the Desktop Calculators section. Even some interesting trivia is given on the site, like why calculator key pads are numbered in a different order than telephone key pads. There are many links to other sites and articles related to classic calculator technology.

  14. Application of Core Dynamics Modeling to Core-Mantle Interactions

    NASA Technical Reports Server (NTRS)

    Kuang, Weijia

    2003-01-01

    Observations have demonstrated that length of day (LOD) variation on decadal time scales results from exchange of axial angular momentum between the solid mantle and the core. There are in general four core-mantle interaction mechanisms that couple the core and the mantle. Of which, three have been suggested likely the dominant coupling mechanism for the decadal core-mantle angular momentum exchange, namely, gravitational core-mantle coupling arising from density anomalies in the mantle and in the core (including the inner core), the electromagnetic coupling arising from Lorentz force in the electrically conducting lower mantle (e.g. D-layer), and the topographic coupling arising from non-hydrostatic pressure acting on the core-mantle boundary (CMB) topography. In the past decades, most effort has been on estimating the coupling torques from surface geomagnetic observations (kinematic approach), which has provided insights on the core dynamical processes. In the meantime, it also creates questions and concerns on approximations in the studies that may invalidate the corresponding conclusions. The most serious problem is perhaps the approximations that are inconsistent with dynamical processes in the core, such as inconsistencies between the core surface flow beneath the CMB and the CMB topography, and that between the D-layer electric conductivity and the approximations on toroidal field at the CMB. These inconsistencies can only be addressed with numerical core dynamics modeling. In the past few years, we applied our MoSST (Modular, Scalable, Self-consistent and Three-dimensional) core dynamics model to study core-mantle interactions together with geodynamo simulation, aiming at assessing the effect of the dynamical inconsistencies in the kinematic studies on core-mantle coupling torques. We focus on topographic and electromagnetic core-mantle couplings and find that, for the topographic coupling, the consistency between the core flow and the CMB topography is critical for correct evaluation of the coupling torque.

  15. RADIATION MAGNETOHYDRODYNAMIC SIMULATIONS OF PROTOSTELLAR COLLAPSE: PROTOSTELLAR CORE FORMATION

    SciTech Connect

    Tomida, Kengo [Department of Astrophysical Sciences, Princeton University, Princeton, NJ 08544 (United States)] [Department of Astrophysical Sciences, Princeton University, Princeton, NJ 08544 (United States); Tomisaka, Kohji [Department of Astronomical Science, The Graduate University for Advanced Studies (SOKENDAI), Osawa, Mitaka, Tokyo 181-8588 (Japan)] [Department of Astronomical Science, The Graduate University for Advanced Studies (SOKENDAI), Osawa, Mitaka, Tokyo 181-8588 (Japan); Matsumoto, Tomoaki [Faculty of Humanity and Environment, Hosei University, Fujimi, Chiyoda-ku, Tokyo 102-8160 (Japan)] [Faculty of Humanity and Environment, Hosei University, Fujimi, Chiyoda-ku, Tokyo 102-8160 (Japan); Hori, Yasunori; Saigo, Kazuya [National Astronomical Observatory of Japan, Osawa, Mitaka, Tokyo 181-8588 (Japan)] [National Astronomical Observatory of Japan, Osawa, Mitaka, Tokyo 181-8588 (Japan); Okuzumi, Satoshi [Department of Physics, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8602 (Japan)] [Department of Physics, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8602 (Japan); Machida, Masahiro N., E-mail: tomida@astro.princeton.edu, E-mail: tomisaka@th.nao.ac.jp, E-mail: yasunori.hori@nao.ac.jp, E-mail: saigo.kazuya@nao.ac.jp, E-mail: matsu@hosei.ac.jp, E-mail: okuzumi@nagoya-u.jp, E-mail: machida.masahiro.018@m.kyushu-u.ac.jp [Department of Earth and Planetary Sciences, Faculty of Sciences, Kyushu University, Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan)

    2013-01-20

    We report the first three-dimensional radiation magnetohydrodynamic (RMHD) simulations of protostellar collapse with and without Ohmic dissipation. We take into account many physical processes required to study star formation processes, including a realistic equation of state. We follow the evolution from molecular cloud cores until protostellar cores are formed with sufficiently high resolutions without introducing a sink particle. The physical processes involved in the simulations and adopted numerical methods are described in detail. We can calculate only about one year after the formation of the protostellar cores with our direct three-dimensional RMHD simulations because of the extremely short timescale in the deep interior of the formed protostellar cores, but successfully describe the early phase of star formation processes. The thermal evolution and the structure of the first and second (protostellar) cores are consistent with previous one-dimensional simulations using full radiation transfer, but differ considerably from preceding multi-dimensional studies with the barotropic approximation. The protostellar cores evolve virtually spherically symmetric in the ideal MHD models because of efficient angular momentum transport by magnetic fields, but Ohmic dissipation enables the formation of the circumstellar disks in the vicinity of the protostellar cores as in previous MHD studies with the barotropic approximation. The formed disks are still small (less than 0.35 AU) because we simulate only the earliest evolution. We also confirm that two different types of outflows are naturally launched by magnetic fields from the first cores and protostellar cores in the resistive MHD models.

  16. 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.

  17. A 100 MWe advanced sodium-cooled fast reactor core concept

    SciTech Connect

    Kim, T. K.; Grandy, C.; Hill, R. N. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2012-07-01

    An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

  18. Ab initio no core full configuration approach for light nuclei

    NASA Astrophysics Data System (ADS)

    Kim, Youngman; Shin, Ik Jae; Maris, Pieter; Vary, James P.; Forssén, Christian; Rotureau, Jimmy

    2014-07-01

    Comprehensive understanding of the structure and reactions of light nuclei poses theoretical and computational challenges. Still, a number of ab initio approaches have been developed to calculate the properties of atomic nuclei using fundamental interactions among nucleons. Among them, we work with the ab initio no core full configuration (NCFC) method and ab initio no core Gamow Shell Model (GSM). We first review these approaches and present some recent results.

  19. The gravitational-wave signature of core-collapse supernovae

    Microsoft Academic Search

    Christian D. Ott; Niels Bohr

    2009-01-01

    We review the ensemble of anticipated gravitational-wave (GW) emission processes in stellar core collapse and postbounce core-collapse supernova evolution. We discuss recent progress in the modeling of these processes and summarize most recent GW signal estimates. In addition, we present new results on the GW emission from postbounce convective overturn and protoneutron star g-mode pulsations based on axisymmetric radiation-hydrodynamic calculations.

  20. Core loss in buried magnet permanent magnet synchronous motors

    Microsoft Academic Search

    R. Schifer; T.A. Lipo

    1989-01-01

    The steady-state core-loss characteristics of buried-magnet synchronous motors operating from a sinusoidal constant frequency voltage supply are investigated. Measured and calculated core loss, with constant shaft load, is shown to increase with decreasing terminal voltage due to an increase in armature reaction-induced stator flux-density time harmonics. Finite-element modeling is used to show that the additional loss due to the time-harmonic

  1. Ab initio no core full configuration approach for light nuclei

    NASA Astrophysics Data System (ADS)

    Kim, Youngman; Shin, Ik Jae; Maris, Pieter; Vary, James P.; Forssén, Christian; Rotureau, Jimmy

    2015-10-01

    Comprehensive understanding of the structure and reactions of light nuclei poses theoretical and computational challenges. Still, a number of ab initio approaches have been developed to calculate the properties of atomic nuclei using fundamental interactions among nucleons. Among them, we work with the ab initio no core full configuration (NCFC) method and ab initio no core Gamow Shell Model (GSM). We first review these approaches and present some recent results.

  2. Solving the Heterogeneous VHTR Core with Efficient Grid Computing

    NASA Astrophysics Data System (ADS)

    Connolly, Kevin John; Rahnema, Farzad

    2014-06-01

    This paper uses the coarse mesh transport method COMET to solve the eigenvalue and pin fission density distribution of the Very High Temperature Reactor (VHTR). It does this using the Boltzmann transport equation without such low-order approximations as diffusion, and it does not simplify the reactor core problem through homogenization techniques. This method is chosen as it makes highly efficient use of grid computing resources: it conducts a series of calculations at the block level using Monte Carlo to model the explicit geometry within the core without approximation, and compiles a compendium of data with the solution set. From there, it is able to solve the desired core configuration on a single processor in a fraction of the time necessary for whole-core deterministic or stochastic transport calculations. Thus, the method supplies a solution which has the accuracy of a whole-core Monte Carlo solution via the computing power available to the user. The core solved herein, the VHTR, was chosen due to its complexity. With a high level of detailed heterogeneity present from the core level to the pin level, and with asymmetric blocks and control material present outside of the fueled region of the core, this reactor geometry creates problems for methods which rely on homogenization or diffusion methods. Even transport methods find it challenging to solve. As it is desirable to reduce the number of assumptions necessary for a whole core calculation, this choice of reactor and solution method combination is an appropriate choice for a demonstration on an efficient use of grid computing.

  3. Ditch the Calculators.

    ERIC Educational Resources Information Center

    Hunsaker, Diane

    1997-01-01

    Discusses the overuse of calculators in mathematics classrooms. Argues that the sole purpose of teaching mathematics is for thinking and discipline and these goals cannot be obtained when using calculators. (ASK)

  4. Personal Finance Calculations.

    ERIC Educational Resources Information Center

    Argo, Mark

    1982-01-01

    Contains explanations and examples of mathematical calculations for a secondary level course on personal finance. How to calculate total monetary cost of an item, monthly payments, different types of interest, annual percentage rates, and unit pricing is explained. (RM)

  5. Calculators in the Classroom?

    ERIC Educational Resources Information Center

    Royce, George; Shank, James

    1977-01-01

    Reports results of a student attitude survey among junior high school students who had been allowed to utilize classroom calculators to check results of mathematical computations. Students displayed a significant preference for using calculators in the classroom. (SL)

  6. A seismologically consistent compositional model of Earth's core.

    PubMed

    Badro, James; Côté, Alexander S; Brodholt, John P

    2014-05-27

    Earth's core is less dense than iron, and therefore it must contain "light elements," such as S, Si, O, or C. We use ab initio molecular dynamics to calculate the density and bulk sound velocity in liquid metal alloys at the pressure and temperature conditions of Earth's outer core. We compare the velocity and density for any composition in the (Fe-Ni, C, O, Si, S) system to radial seismological models and find a range of compositional models that fit the seismological data. We find no oxygen-free composition that fits the seismological data, and therefore our results indicate that oxygen is always required in the outer core. An oxygen-rich core is a strong indication of high-pressure and high-temperature conditions of core differentiation in a deep magma ocean with an FeO concentration (oxygen fugacity) higher than that of the present-day mantle. PMID:24821817

  7. Ice Chemistry in Starless Molecular Cores

    NASA Astrophysics Data System (ADS)

    Kalv?ns, J.

    2015-06-01

    Starless molecular cores are natural laboratories for interstellar molecular chemistry research. The chemistry of ices in such objects was investigated with a three-phase (gas, surface, and mantle) model. We considered the center part of five starless cores, with their physical conditions derived from observations. The ice chemistry of oxygen, nitrogen, sulfur, and complex organic molecules (COMs) was analyzed. We found that an ice-depth dimension, measured, e.g., in monolayers, is essential for modeling of chemistry in interstellar ices. Particularly, the H2O:CO:CO2:N2:NH3 ice abundance ratio regulates the production and destruction of minor species. It is suggested that photodesorption during the core-collapse period is responsible for the high abundance of interstellar H2O2 and O2H and other species synthesized on the surface. The calculated abundances of COMs in ice were compared to observed gas-phase values. Smaller activation barriers for CO and H2CO hydrogenation may help explain the production of a number of COMs. The observed abundance of methyl formate HCOOCH3 could be reproduced with a 1 kyr, 20 K temperature spike. Possible desorption mechanisms, relevant for COMs, are gas turbulence (ice exposure to interstellar photons) or a weak shock within the cloud core (grain collisions). To reproduce the observed COM abundances with the present 0D model, 1%–10% of ice mass needs to be sublimated. We estimate that the lifetime for starless cores likely does not exceed 1 Myr. Taurus cores are likely to be younger than their counterparts in most other clouds.

  8. The Core Velocity Dispersion (CVD) for Taurus Dense Core Clusters

    NASA Astrophysics Data System (ADS)

    Li, Di; Qian, L.

    2012-05-01

    We define a statistical measurement--Core Velocity Dispersion (CVD) for the dynamics of a dense core cluster. To obtain a well defined CVD requires a spectroscopic dense core sample located in a contiguous region. We measure CVD for Taurus cores utilizing the 100 d^2 13CO map corrected for depletion. The Taurus CVD has the same power law as that of the Larson's law. No sign of additional energy input is seen.

  9. Core-Valence Auger Spectra of the Simple Metals.

    NASA Astrophysics Data System (ADS)

    Davis, Linda Rae

    A comprehensive study of the core-valence-valence (CVV) Auger line shapes for the simple metals, Li, Be, Na, Mg, and Al, is presented. Calculations of these line shapes are performed with a static many-body model which treats the (static) screening of a core hole by conduction electrons with a phenomenological potential. Our model is applicable only to the simple metals, because free-electron theory is used to describe the conduction electrons. The model is implemented with a finite-particle, determinantal approach, and many-body Auger matrix elements are calculated explicitly using a local approximation. The many-body effects which arise from the core-conduction interaction are treated exactly within the model. The dependence of the model line shapes (i) on parameters specifying the metal and core level, (ii) on approximation of the model, and (iii) on the many-body effects produced by the core -conduction interaction are examined in detail. Also, model calculations of the simple-metal CVV spectra for which data exist are presented. Empirical fits of the model calculations to reported data are achieved for the Li(KVV), Be(KVV), Mg(L(,2,3)VV), and Al(L(,2,3)VV) spectra. In all of these cases, the screening of the core hole by conduction electrons is found to have both s-like and p-like character. One -electron and static many-body calculations have also been performed by other authors for some of the simple-metal spectra. Our model and results are compared to those from the other calculations. Although one-electron theory has reproduced several of the reported simple-metal line shapes, our calculations are the only many-body calculations to produce line shapes in agreement with data. Model calculations for the final reported spectrum, the Mg(KVV), are not presented, because the Mg K core hole is not sufficiently long-lived for our model to be valid. Our expectations concerning the unreported Na(L(,2,3)VV) spectrum are discussed. From our empirical fits and model study, we conclude that the many-body effects which are introduced by the core-conduction interaction, although weak, significantly influence the CVV line shapes of the simple metals. This conclusion is suggestive of the inadequacy of the one-electron theory of CVV line shapes, even in cases of simple-metal spectra.

  10. Calculation of the spectrum of the superheavy element Z=120

    E-print Network

    Dinh, T H; Flambaum, V V; Ginges, J S M

    2008-01-01

    High-precision calculations of the energy levels of the superheavy element Z=120 are presented. The relativistic Hartree-Fock and configuration interaction techniques are employed. The correlations between core and valence electrons are treated by means of the correlation potential method and many-body perturbation theory. Similar calculations for barium and radium are used to gauge the accuracy of the calculations and to improve the ab initio results.

  11. Calculator-Active Materials.

    ERIC Educational Resources Information Center

    Crow, Tracy, Ed.; Harris, Julia, Ed.

    1997-01-01

    This journal contains brief descriptions of calculator-active materials that were found using Resource Finder, the searchable online catalog of curriculum resources from the Eisenhower National Clearinghouse (ENC). It features both the calculators themselves and the activity books that are used with them. Among the calculators included are those…

  12. Calculating a Mineral's Density

    NSDL National Science Digital Library

    Andrea Distelhurst

    2011-10-05

    Students will use the Density=Mass/Volume formula to calculate the density of an unknown mineral. By using water displacement and a triple beam balance students will collect measurements of volume and mass for an unknown mineral. With this data, they will calculate the mineral's density then identify the mineral based on calculated density.

  13. Ecological Footprint Calculators

    NSDL National Science Digital Library

    EcoBusinessLinks

    This website contains interactive calculators for determining various environmental impacts. The site includes more than 15 different calculators to determine greenhouse gas emissions, ecological footprints, electricity pollution, air travel pollution, commuting costs, appliance costs, pollution prevention and more. These calculators can be used for computer-based classroom activities or to enable students to see which types of activities have the greatest environmental impact.

  14. Calculators in the Classroom

    ERIC Educational Resources Information Center

    Pendleton, Deedee

    1975-01-01

    Presents the pro and con of the use of calculators in the classroom. Some feel that calculators make learning mathematics more fun and when used for creative problem solving provide student motivation. Others feel that dependence on the calculators will result in students unable to do simple mathematics on paper. (GS)

  15. All-University Core Curriculum All-University Core Curriculum

    E-print Network

    Stephens, Graeme L.

    All-University Core Curriculum _______________ 2.3 Page 1 All-University Core Curriculum Office-UNIVERSITY CORE CURRICULUM (AUCC) All Colorado State University students share a learning experience in common. Mathematics1 3 2. Advanced Writing (3 credits)2, 3 3. Foundations and Perspectives (22 credits) A. Biological

  16. Nutrient Composition of the Peanut Core of the Core Collection

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Peanuts from the Core Collection designated as the Core of the Core Collection were grown in Tifton, GA in 2005. Amino acids, folic acid and total oil content were determined on the whole seed. Amino acid concentrations were generally close to commonly reported values. Folic acid concentration var...

  17. Prediction of dislocation cores in aluminum from density functional theory.

    PubMed

    Woodward, C; Trinkle, D R; Hector, L G; Olmsted, D L

    2008-02-01

    The strain field of isolated screw and edge dislocation cores in aluminum are calculated using density-functional theory and a flexible boundary condition method. Nye tensor density contours and differential displacement fields are used to accurately bound Shockley partial separation distances. Our results of 5-7.5 A (screw) and 7.0-9.5 A (edge) eliminate uncertainties resulting from the wide range of previous results based on Peierls-Nabarro and atomistic methods. Favorable agreement of the predicted cores with limited experimental measurements demonstrates the need for quantum mechanical treatment of dislocation cores. PMID:18352300

  18. The structure of the chromatin core particle in solution.

    PubMed Central

    Pardon, J F; Worcester, D L; Wooley, J C; Cotter, R I; Lilley, D M; Richards, R M

    1977-01-01

    The shape and size of the nucleosomal core particle from chromatin has been examined by analysis of neutron and X-ray scattering data from dilute solutions. Calculations of scattering for many different models have been made and only one model was able to account for both the X-ray and neutron profiles. This model is an oblate structure with height about 50A and diameter 110A. The DNA is mainly confined to two annuli located at the top and bottom respectively of the core particle positioned on the outside of a compact protein core which has a height of about 40A and diameter about 73A. PMID:561952

  19. First-principles investigation of pentagonal and hexagonal core-shell silicon nanowires with various core compositions

    Microsoft Academic Search

    C. Berkdemir; O. Gülseren

    2009-01-01

    Properties of various core-shell silicon nanowires are investigated by extensive first-principles calculations on the geometric optimization as well as electronic band structures of the nanowires by using pseudopotential plane-wave method based on the density-functional theory. We show that different geometrical structures of silicon nanowires with various core compositions, formed by stacking of atomic polygons with pentagonal or hexagonal cross sections

  20. Constraint on the 1D earth model near core-mantle boundary by free core nutation

    NASA Astrophysics Data System (ADS)

    Huang, Chengli; Zhang, Mian

    2015-04-01

    Free core nutation (FCN) is a normal mode of the rotating earth with fluid outer core (FOC). Its period depends on the physics of the mantle and FOC, especially the parameters near core-mantle boundary (CMB), like the density and elastic (Lame) parameters. FCN period can be determined very accurately by VLBI and superconductive tidal gravimetry, but the theoretical calculation results of FCN period from traditional approaches and 1D earth model (like PREM) deviate significantly from the accurate observation. Meanwhile, the influence of the uncertainty of a given earth model on nutation has never been studied before. In this work, a numerical experiment is presented to check this problem, and we want to see whether FCN can provide a constraint on the construction of a 1D earth model, especially on the gradient of material density near CMB.

  1. Geomagnetism of earth's core

    NASA Technical Reports Server (NTRS)

    Benton, E. R.

    1983-01-01

    Instrumentation, analytical methods, and research goals for understanding the behavior and source of geophysical magnetism are reviewed. Magsat, launched in 1979, collected global magnetometer data and identified the main terrestrial magnetic fields. The data has been treated by representing the curl-free field in terms of a scalar potential which is decomposed into a truncated series of spherical harmonics. Solutions to the Laplace equation then extend the field upward or downward from the measurement level through intervening spaces with no source. Further research is necessary on the interaction between harmonics of various spatial scales. Attempts are also being made to analytically model the main field and its secular variation at the core-mantle boundary. Work is also being done on characterizing the core structure, composition, thermodynamics, energetics, and formation, as well as designing a new Magsat or a tethered satellite to be flown on the Shuttle.

  2. Neutronic analysis of boiling water reactor in-core detector noise

    Microsoft Academic Search

    H. S. Cheng; D. J. Diamond

    1979-01-01

    The response of boiling water reactor in-core detectors undergoing vibration has been calculated. A neutronic model based on calculating the fission activity at a detector position in a planar multibundle environment was employed. The model used eight energy groups and two-dimensional Cartesian geometry in a discrete-ordinates transport approximation. The in-core detector responses due to various detector displacements were calculated as

  3. Measuring Core Inflation

    Microsoft Academic Search

    Michael F. Bryan; Stephen G. Cecchetti

    1993-01-01

    In this paper, we investigate the use of limited-information estimators as measures of core inflation. Employing a model of asymmetric supply disturbances, with costly price adjustment, we show how the observed skewness in the cross-sectional distribution of inflation can cause substantial noise in the aggregate price index at high frequencies. The model suggests that limited-influence estimators, such as the median

  4. Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Albright, Dennis; Butler, Carey; West, Nicole; Cole, John W. (Technical Monitor)

    2002-01-01

    Institute for Scientific Research, Inc. (ISR) research program consist of: 1.Study core physics by adapting existing codes: MCNP4C - Monte Carlo code; COMBINE/VENTURE - diffusion theory; SCALE4 - Monte Carlo, with many utility codes. 2. Determine feasibility and study major design parameters: fuel selection, temperature and reflector sizing. 3. Study reactor kinetics: develop QCALC1 to model point kinetics; study dynamic behavior of the power release.

  5. Banded electromagnetic stator core

    DOEpatents

    Fanning, Alan W. (San Jose, CA); Gonzales, Aaron A. (San Jose, CA); Patel, Mahadeo R. (San Jose, CA); Olich, Eugene E. (Aptos, CA)

    1996-01-01

    A stator core for an electromagnetic pump includes a plurality of circumferentially adjoining groups of flat laminations disposed about a common centerline axis and collectively defining a central bore and a discontinuous outer perimeter, with adjacent groups diverging radially outwardly to form V-shaped gaps. An annular band surrounds the groups and is predeterminedly tensioned to clamp together the laminations, and has a predetermined flexibility in a radial direction to form substantially straight bridge sections between the adjacent groups.

  6. Banded electromagnetic stator core

    DOEpatents

    Fanning, Alan W. (San Jose, CA); Gonzales, Aaron A. (San Jose, CA); Patel, Mahadeo R. (San Jose, CA); Olich, Eugene E. (Aptos, CA)

    1994-01-01

    A stator core for an electromagnetic pump includes a plurality of circumferentially adjoining groups of flat laminations disposed about a common centerline axis and collectively defining a central bore and a discontinuous outer perimeter, with adjacent groups diverging radially outwardly to form V-shaped gaps. An annular band surrounds the groups and is predeterminedly tensioned to clamp together the laminations, and has a predetermined flexibility in a radial direction to form substantially straight bridge sections between the adjacent groups.

  7. Banded electromagnetic stator core

    DOEpatents

    Fanning, A.W.; Gonzales, A.A.; Patel, M.R.; Olich, E.E.

    1994-04-05

    A stator core for an electromagnetic pump includes a plurality of circumferentially adjoining groups of flat laminations disposed about a common centerline axis and collectively defining a central bore and a discontinuous outer perimeter, with adjacent groups diverging radially outwardly to form V-shaped gaps. An annular band surrounds the groups and is predeterminedly tensioned to clamp together the laminations, and has a predetermined flexibility in a radial direction to form substantially straight bridge sections between the adjacent groups. 5 figures.

  8. Central core disease

    PubMed Central

    Jungbluth, Heinz

    2007-01-01

    Central core disease (CCD) is an inherited neuromuscular disorder characterised by central cores on muscle biopsy and clinical features of a congenital myopathy. Prevalence is unknown but the condition is probably more common than other congenital myopathies. CCD typically presents in infancy with hypotonia and motor developmental delay and is characterized by predominantly proximal weakness pronounced in the hip girdle; orthopaedic complications are common and malignant hyperthermia susceptibility (MHS) is a frequent complication. CCD and MHS are allelic conditions both due to (predominantly dominant) mutations in the skeletal muscle ryanodine receptor (RYR1) gene, encoding the principal skeletal muscle sarcoplasmic reticulum calcium release channel (RyR1). Altered excitability and/or changes in calcium homeostasis within muscle cells due to mutation-induced conformational changes of the RyR protein are considered the main pathogenetic mechanism(s). The diagnosis of CCD is based on the presence of suggestive clinical features and central cores on muscle biopsy; muscle MRI may show a characteristic pattern of selective muscle involvement and aid the diagnosis in cases with equivocal histopathological findings. Mutational analysis of the RYR1 gene may provide genetic confirmation of the diagnosis. Management is mainly supportive and has to anticipate susceptibility to potentially life-threatening reactions to general anaesthesia. Further evaluation of the underlying molecular mechanisms may provide the basis for future rational pharmacological treatment. In the majority of patients, weakness is static or only slowly progressive, with a favourable long-term outcome. PMID:17504518

  9. Shape Sensing Using a Multi-Core Optical Fiber Having an Arbitrary Initial Shape in the Presence of Extrinsic Forces

    NASA Technical Reports Server (NTRS)

    Rogge, Matthew D. (Inventor); Moore, Jason P. (Inventor)

    2014-01-01

    Shape of a multi-core optical fiber is determined by positioning the fiber in an arbitrary initial shape and measuring strain over the fiber's length using strain sensors. A three-coordinate p-vector is defined for each core as a function of the distance of the corresponding cores from a center point of the fiber and a bending angle of the cores. The method includes calculating, via a controller, an applied strain value of the fiber using the p-vector and the measured strain for each core, and calculating strain due to bending as a function of the measured and the applied strain values. Additionally, an apparent local curvature vector is defined for each core as a function of the calculated strain due to bending. Curvature and bend direction are calculated using the apparent local curvature vector, and fiber shape is determined via the controller using the calculated curvature and bend direction.

  10. Towards Core Collapse Supernova Models for Nucleosynthesis

    NASA Astrophysics Data System (ADS)

    Casanova, Jordi; Frohlich, C.; Perego, A.; Hempel, M.; Liebendoerfer, M.

    2013-01-01

    Massive stars (M >= 8-10 Msun) undergo core collapse at the end of their life and explode with a kinetic energy release of 10^51 ergs. While the detailed explosion mechanism is still under investigation using (computationally expensive) multi-dimensional simulations, accurate and efficient core collapse supernova simulations are needed for detailed nucleosynthesis predictions. Detailed nucleosynthesis calculations are required to explain the observed abundances in metal-poor stars, and will provide supernova yields for galactic chemical evolution studies. We are modeling the core collapse, bounce and subsequent explosion of massive stars assuming spherical symmetry by using the code Agile-IDSA. We trigger the explosion by depositing a fraction of the neutrino energy into the heating region, so that the stalled shock is revived. The code Agile-IDSA is based on the hydrodynamics code Agile and employs the isotropic diffusion source approximation (IDSA) by Liebendoerfer et al (2009) for the neutrino transport. This combination provides a computationally efficient tool to study aspects of core-collapse supernovae where solving the full Boltzmann transport equations are computationally less efficient. We will present results from recent simulations implementing a modern supernova equation of state which includes inhomogeneous nuclear matter. We will discuss the feasibility of long term self-consistent explosions which cover the entire duration of nucleosynthesis. We will also study the influence of the progenitor model on the explosion properties.

  11. Effect of Rotation in Cloud Core Collapse

    NASA Astrophysics Data System (ADS)

    Tsuribe, T.

    The collapse of rotating clouds is investigated.At first, isothermal collapse of an initially uniform-density, uniform-rotating, molecular cloud core with pressure and self-gravity is investigated to determine the conditions under which a cloud is unstable to fragmentation. A semianalytic model for the collapse of rotating spheroids is developed with the method of characteristics for inwardly propagating rarefaction waves. Three-dimensional self-gravitating hydrodynamical calculations are performed for the initially uniform-density rigid-rotating sphere. Both investigations show that the criterion for fragmentation is modified from the one in the literature if the property of the non-homologous collapse is taken into account. It is shown that the central flatness, that is, the axial ratio of the isodensity contour in the central region, is a good indicator for the fate of the cloud. We derive the criterion for the fragmentation considering the evolution of the flatness of the central core. If the central flatness becomes greater than the critical value ˜ 4?, a collapsing cloud with moderate perturbations is unstable for fragmentation, while if the central flatness stays smaller than the critical value, it does not fragment at least before adiabatic core formation. Warm clouds (?0 ? 0.5) are not expected to fragment before adiabatic core formation almost independent of the initial rotation (?0) and the properties of the initial perturbation. The effect of the initial density central concentration is also investigated. If it exists, distortion or flattening of a cloud core is suppressed even if ?0 ? 0.5 in small rotation cases due to stronger nonhomologous property of the collapse. We conclude that the binary fragmentation is difficult during isothermal stage if a core collapse had started from near spherical configurations with moderate thermal energy or small rotation. We suggest that the close binary fragmentation may be possible in the nonisothermal stage by rapid growth of a nonspherical first core. Second, to mimic fragmentation processes of primordial clouds, the equation of state is approximated by a simple polytropic relation with ? ˜ 1.1. A series of numerical and semianalytical calculations of the rotating collapse of an initially spherical cloud shows a criterion for fragmentation of rotating polytropic cloud cores with ?=1.1. Fragmentation during core collapse is not expected to take place if the cloud thermal energy is greater than 0.3 times its gravitational energy at the initial stage of runaway collapse. The collapse of the central small core will not be halted by centrifugal force since a nonaxisymmetric waves will appear and the flow will converge to a self-similar flow until ? exceeds 4/3. Finally, we take into account the detailed non-equilibrium chemical reactions for primordial gas that consists of pure hydrogen. The parameters of the collapse and the condition of the fragmentation are compared with those of isothermal clouds. It is shown that the geometrical flatness of the central region of the disc is a good indicator for predicting whether the clouds fragment or not. If the flatness is greater than the critical value, ˜ 4?, a cloud fragments into filaments and blobs. On the other hand, if the flatness is smaller than the critical value, fragmentation is not expected before the central core formation even if the cooling is efficient and the total mass becomes much greater than the local Jeans mass at the center.

  12. (w13) Institutional Core Management

    PubMed Central

    Turpen, P.; Farber, G.K.; Mische, S.; Alexander, P.; Auger, J.; Meyn, S.

    2011-01-01

    This workshop session will focus on issues related to Institutional Core Management, in response to the national conversation evolving around research core facility issues and management. The workshop will be formatted as an experts' panel; each participant currently plays an important role in supporting and developing research core resources at an institutional level. Some of the topics to be discussed include: (1) Core Consolidation — one size fits all? (2) Bottom-up vs. top-down management, advantages and disadvantages of centrally managed cores. (3) Performance metrics and impacts on professional development, core infrastructure support and improved operations. (4) Impacts of NIH-NCRR programs on improving access to research resources, including core facilities. We also plan to highlight the new Core Administrators Network Coordinating (CAN). In response to an emerging trend to centralize the oversight of research core facilities, ABRF has fostered development of this network and a new committee: the Core Administrators Network-Coordinating Committee (CAN-CC). The committee seeks input and participation from scientists, administrators and others with an interest in issues related to the administration of research core facilities which, by the nature of their service role, must interface with multiple constituencies within a research enterprise. Today many institutions have established administrative positions designed to assist core facilities with management of economic, regulatory and performance issues. In order to facilitate greater interaction between and among core scientists and administrators, the mission of the CAN-CC is to contribute to the common interests of core administrators, and promote interactions with core scientists in a collegial and productive manner. The specific goals of the Core Administrators Network Coordinating Committee (CANCC) are: to identify and reach out to our target community; provide opportunities for networking; and assess goals for program focus and development.

  13. Models of the earth's core

    Microsoft Academic Search

    D. J. Stevenson

    1981-01-01

    The combination of seismology, high pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to strong constraints on core models. The synthesis presented here is devoted to the defense of the following properties: (1) core formation was contemporaneous with earth accretion; (2) the outer, liquid core is predominately iron but cannot be purely iron;

  14. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, 1 Plant Road, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  15. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, 1 Plant Road, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ?50% content of low-power blanket bundles may require power de-rating (?58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  16. hp calculators HP 50g Calculator Modes and Customization

    E-print Network

    Vetter, Frederick J.

    hp calculators HP 50g Calculator Modes and Customization Choosing how to use your Calculator and Customization #12;hp calculators HP 50g Calculator Modes and Customization hp calculators - 2 - HP 50g Calculator Modes and Customization Choosing how to Use your Calculator The HP 50g gives you many ways

  17. Do Super-Earths Have Convective Cores?

    NASA Astrophysics Data System (ADS)

    Gold, M. W.; Leitner, J. J.; Firneis, M. G.

    2012-04-01

    The processes within the Earth responsible for the generation of the magnetic field are known fairly well by now and highly sophisticated 3D dynamo models are already capable of simulating magnetic field reversals (Glatzmaier and Roberts, 1995). But do the assumptions about the Earth's interior, like vigorous convection in the outer core, also hold for Super-Earths, terrestrial exoplanets more massive than the Earth? To shed light on this matter, as a first step an interior structure model (e.g., Wagner et al., 2011) has to be applied in order to determine density, pressure and temperature gradients of the planetary structure under consideration, in conjunction with an adequate equation of state that accounts for high-pressure conditions in the deeper regions of the planet. By coupling a structure model with a parameterized convection approach for the mantle (e.g., Nimmo et al., 2004), the heat flow across the core-mantle boundary can be determined. This heat flow is crucial in assessing the ability of the planet's core for vigorous convection, one prerequisite for dynamo action. If the heat flow is lower than the heat conducted down the adiabat, conduction will dominate, and thus convection will be restrained. By using the model curves calculated by Wagner et al. (2011), and employing the parameterized approach demonstrated by Nimmo et al. (2004), the CMB heat flow has been estimated for the case of 5 ME and 10 ME planets, assuming 205 ppm 40K in the core . It turns out that the heat conducted down the adiabat Qk exceeds the heat flow across the core-mantle boundary QC in both cases: Qk=33.37 TW and 93.6 TW for 5 ME and 10 ME planets, respectively, using the results for the Keane-EOS from Wagner et al. (2011), while QC=5.61 TW and 12.13 TW, respectively. The difficulties involved with the parameterized approach (Labrosse, 2003) and the poorly constrained values of core quantities, like the thermal expansivity, or thermal conductivity at the base of the mantle, can lead to substantial inaccuracies in the estimated heat flows, but it seems to be very likely that the qualitative trend is correct: the increasing dominance of conduction in planetary cores and mantles with increasing planetary mass. Although radiogenic heat has been included in the energy budget of the core in this preliminary model, its influence on viscosity hasn't been scrutinized so far. Moreover, the possibility of compositional convection at pressures prevalent in massive terrestrial exoplanets should be subject of further studies.

  18. Characteristics of and corrections for core shortening in unconsolidated sediments

    USGS Publications Warehouse

    Morton, Robert A.; White, William A.

    1997-01-01

    Thinning, bypassing, and compaction of shallow unconsolidated sediments during manual coring or vibracoring operations probably cause more sediment deformation and greater stratigraphic displacement than is commonly reported in the wetland literature. We measured core shortening in open-barrel cores from fluvial wetlands, lagoonal flats, and marshes to document the magnitude and characteristics of shortening where sediments may be stiff and require extra mechanical effort to recover a sufficient length of sample for analysis. Results of those measurements indicate that thinning or non- recovery of discrete sediment intervals can range from 0 to 67 percent and cumulative core shortening can be as much as 30 percent even for cores less than one meter long. Detailed open-barrel measurements also show that core shortening is not uniformly distributed throughout the depth of penetration as is often assumed. Analytical data derived from shortened cores can only be properly interpreted if patterns of shortening are established and incorporated into the analysis. Minor artificial displacement of sediment depths can alter plots of physico-chemical parameters and can significantly influence calculated rates of sedimentation and other depth-dependent statistical relationships. This study (1) demonstrates how plots of interval shortening and cumulative shortening can be used to characterize the distribution of shortening at depth and (2) presents a simple equation for stratigraphic restoration so that core observations and analyses are corrected to their original depths.

  19. Gas Hydrate-Sediment Morphologies Revealed by Pressure Core Analysis

    NASA Astrophysics Data System (ADS)

    Holland, M.; Schultheiss, P.; Roberts, J.; Druce, M.

    2006-12-01

    Analysis of HYACINTH pressure cores collected on IODP Expedition 311 and NGHP Expedition 1 showed gas hydrate layers, lenses, and veins contained in fine-grained sediments as well as gas hydrate contained in coarse-grained layers. Pressure cores were recovered from sediments on the Cascadia Margin off the North American West Coast and in the Krishna-Godavari Basin in the Western Bay of Bengal in water depths of 800- 1400 meters. Recovered cores were transferred to laboratory chambers without loss of pressure and nondestructive measurements were made at in situ pressures and controlled temperatures. Gamma density, P-wave velocity, and X-ray images showed evidence of grain-displacing and pore-filling gas hydrate in the cores. Data highlights include X-ray images of fine-grained sediment cores showing wispy subvertical veins of gas hydrate and P-wave velocity excursions corresponding to grain-displacing layers and pore-filling layers of gas hydrate. Most cores were subjected to controlled depressurization experiments, where expelled gas was collected, analyzed for composition, and used to calculate gas hydrate saturation within the core. Selected cores were stored under pressure for postcruise analysis and subsampling.

  20. Flux harmonics in large SFR cores in relation with core characteristics such as power peaks

    SciTech Connect

    Rimpault, G.; Buiron, L.; Fontaine, B.; Sciora, P.; Tommasi, J. [CEA, DEN, DER, SPRC Cadarache, F-13108 Saint Paul-lez-Durance (France)

    2013-07-01

    Designing future Sodium Fast Reactors (SFR) requires enhancing their operational performance and reducing the probability to go into core disruption. As a consequence of these constraints, these novel reactors exhibit rather unusual features compared to past designs. The cores are much larger with rather flat shape. The consequences of that shape on the core characteristics deserve to be studied. The approach taken in this paper is to calculate the eigenvalue associated to the first harmonic and its associated flux. It is demonstrated that these values are linked to some core features, in particular, those sensitive to spatial effects such as power peaks induced by the movement of control rods. The uncertainty associated to these characteristics is being tentatively studied and guidelines for further studied are being identified. In the development strategy of these new SFR designs, a first demonstration plant of limited installed power (around 1500 MWth) will have to be built first. Identifying the possibility of going later to higher power plants (around 3600 MWth) without facing new challenges is an important criterion for designing such a plant. That strategy is being studied, in this paper, focusing on some rather frequent initiator such as the inadvertent control rod withdrawal for different core sizes with the help of the perturbation theory and the flux harmonics. (authors)

  1. Thermochemical Evolution of Earth's Core with Magnesium Precipitation

    NASA Astrophysics Data System (ADS)

    O'Rourke, J. G.; Stevenson, D. J.

    2014-12-01

    Vigorous convection within Earth's outer core drives a dynamo that has sustained a global magnetic field for at least 3.5 Gyr. Traditionally, people invoke three energy sources for the dynamo: thermal convection from cooling and freezing, compositional convection from light elements expelled by the growing inner core, and, perhaps, radiogenic heating from potassium-40. New theoretical and experimental work, however, indicates that the thermal and electrical conductivities of the outer core may be as much as three times higher than previously assumed. The implied increase in the adiabatic heat flux casts doubt on the ability of the usual mechanisms to explain the dynamo's longevity. Here, we present a quantitative model of the crystallization of magnesium-bearing minerals from the cooling core—a plausible candidate for the missing power source. Recent diamond-anvil cell experiments suggest that magnesium can partition into core material if thermodynamic equilibrium is achieved at high temperatures (>5000 K). We develop a model for core/mantle differentiation in which most of the core forms from material equilibrated at the base of a magma ocean as Earth slowly grows, but a small portion (~10%) equilibrated at extreme conditions in the aftermath of a giant impact. We calculate the posterior probability distribution for the original concentrations of magnesium and other light elements (chiefly oxygen and silicon) in the core, constrained by partitioning experiments and the observed depletion of siderophile elements in Earth's mantle. We then simulate the thermochemical evolution of cores with plausible compositions and thermal structures from the end of accretion to the present, focusing on the crystallization of a few percent of the initial core as ferropericlase and bridgmanite. Finally, we compute the associated energy release and verify that our final core compositions are consistent with the available seismological data.

  2. Flexure modelling at seamounts with dense cores

    NASA Astrophysics Data System (ADS)

    Kim, Seung-Sep; Wessel, Paul

    2010-08-01

    The lithospheric response to seamounts and ocean islands has been successfully described by deformation of an elastic plate induced by a given volcanic load. If the shape and mass of a seamount are known, the lithospheric flexure due to the seamount is determined by the thickness of an elastic plate, Te, which depends on the load density and the age of the plate at the time of seamount construction. We can thus infer important thermomechanical properties of the lithosphere from Te estimates at seamounts and their correlation with other geophysical inferences, such as cooling of the plate. Whereas the bathymetry (i.e. shape) of a seamount is directly observable, the total mass often requires an assumption of the internal seamount structure. The conventional approach considers the seamount to have a uniform density (e.g. density of the crust). This choice, however, tends to bias the total mass acting on an elastic plate. In this study, we will explore a simple approximation to the seamount's internal structure that considers a dense core and a less dense outer edifice. Although the existence of a core is supported by various gravity and seismic studies, the role of such volcanic cores in flexure modelling has not been fully addressed. Here, we present new analytic solutions for plate flexure due to axisymmetric dense core loads, and use them to examine the effects of dense cores in flexure calculations for a variety of synthetic cases. Comparing analytic solutions with and without a core indicates that the flexure model with uniform density underestimates Te by at least 25 per cent. This bias increases when the uniform density is taken to be equal to the crustal density. We also propose a practical application of the dense core model by constructing a uniform density load of same mass as the dense core load. This approximation allows us to compute the flexural deflection and gravity anomaly of a seamount in the wavenumber domain and minimize the limitations recognized from the analytic tests. Then, the dense core model is applied to predict the lithospheric flexure beneath Howland Island in the Tokelau seamount chain; these results are compared with the predictions of the uniform density model. Based on age dating of Howland and the age of the seafloor, traditional Te versus age curves predict the elastic plate thickness beneath the seamount to be around 20 km, which is comparable to the best dense core model of Te = 26 km. However, the best uniform density model is found at Te = 12 km, which is significantly less than the predicted. From our investigations of synthetic and real seamount cases, we conclude that the dense core model approximates the true mass distribution of a seamount better than the uniform density model. Finally, we suggest that the role of underplating in flexure modelling may need to be reexamined because the dense core model predicts substantially less deflections than the uniform density model without requiring additional buoyancy caused by underplated material.

  3. 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 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  4. Radioactive Decay Calculator

    NSDL National Science Digital Library

    Alan Enns

    This online calculator computes radioactive decay, timed decay, and timed solid disposal for a databank containing 116 isotopes. It also features University of British Columbia disposal limits and a unit converter and date/time calculators. These tools calculate the half-life for selected isotopes; radioactive decay final activity, given the initial activity and decay time; the decay time, given the initial and final activities; and the decay time, given the mass of a solid and the initial activity.

  5. Calculator Java Applet

    NSDL National Science Digital Library

    This applet is a handy scientific calculator with the ability to do unit conversions on the fly. When you click on the link below, it will pop up in its own window so you can continue browsing after loading the calculator. The calculator window is resizable and will pop up to your preferred size after you close it once and come back to this page again.

  6. Transuranic Transmutation and Criticality Calculation Sensitivity to Heterogeneous Lattice Effects - 12391

    SciTech Connect

    Barbaras, Sean A. [United States Military Academy, West Point, New York 10996 (United States); Knight, Travis W. [University of South Carolina, Columbia, South Carolina 29208 (United States)

    2012-07-01

    Using Mixed Oxide (MOX) fuel in traditional Pressurized Water Reactor (PWR) assemblies has been researched at length and has shown to provide the benefit of transmutation and targets the amount and toxicity of high level waste needed to be managed. Advanced MOX concepts using enriched Uranium Dioxide (UO{sub 2}) are required for multiple recycling of plutonium. The use of MOX and ordinary UO{sub 2} fuel in the same assembly as well as unfueled rods and assembly edge effects contrasts with the unit cell computational assumption of a uniform infinite array of rods. While a deterministic method of calculating the Dancoff factor has traditionally been employed in fuel assembly analysis due to the lighter computational and modeling requirements, this research seeks to determine the validity of the uniform, infinite lattice assumption with respect to Dancoff factor and determine the magnitude of the impact of nonuniform lattice effects on fuel assembly criticality calculations as well as transuranic isotope production and transmutation. This research explored the pin-to-pin interaction in a non-uniform lattice of MOX fuel rods and UO{sub 2} fuel rods through the impact of the calculated Dancoff factors from the deterministic method used in SCALE versus the Monte Carlo method used in the code DANCOFF-MC. Using the Monte Carlo method takes into account the non-uniform lattice effects of having neighboring fuel rods with different cross-sectional spectra whereas the Dancoff factor calculated by SCALE assumes a uniform, infinite lattice of one fuel rod type. Differences in eigenvalue calculations as a function of burnup are present between the two methods of Dancoff factor calculation. The percent difference is greatest at low burnup and then becomes smaller throughout the cycle. Differences in the transmutation rate of transuranic isotopes in the MOX fuel are also present between the Dancoff factor calculation methods. The largest difference is in Pu-239, Pu-242, and Am-241 composition whereas U-238, Pu-242, and Pu-238 composition was not changed by taking into account the non-homogenous lattice effects. Heterogeneous lattice effects do change the calculated eigenvalue and transmutation rate in a non-uniform lattice of MOX fuel rods and UO{sub 2} fuel. However, the uncertainty in the ENDF data used by SCALE in these calculations is large enough that the infinite lattice assumption remains valid. (authors)

  7. Core restoration for crown preparation.

    PubMed

    Larson, Thomas D

    2004-01-01

    This article will review the relevant literature fom 1991-2003, a period of time when adhesive resin luting materials became available and luting crowns with zinc phosphate cements decreased. The review wtill look at the principles suggesting when a core should be placed, what core materials function best, preparation design with a core, luting material choice with a core, and results of clinical trials. Amalgam cores are regarded as the strongest material, best able to withstand adverse stress and restore teeth having the greatest loss of tooth structure. Composite resins, whether chemically cured or light cured, reinforced or not, appear best capable of core restorationfor moderately broken down teeth. Glass ionomer materials are considered too weak to withstand stress as a core material, but are recommended as a base material tofill in undercuts and improve the accuracy of impression and fit of a crown. PMID:15554446

  8. 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

    ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium 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.

  9. Combustion and core noise

    NASA Astrophysics Data System (ADS)

    Mahan, J. Robert; Karchmer, Allen

    1991-08-01

    Two types of aircraft power plant are considered: the gas turbine and the reciprocating engine. The engine types considered are: the reciprocating engine, the turbojet engine, the turboprop engine, and the turbofan engine. Combustion noise in gas turbine engines is discussed, and reciprocating-engine combustion noise is also briefly described. The following subject areas are covered: configuration variables, operational variables, characteristics of combustion and core noise, sources of combustion noise, combustion noise theory and comparison with experiment, available prediction methods, diagnostic techniques, measurement techniques, data interpretation, and example applications.

  10. Core 4: Image Bank

    NSDL National Science Digital Library

    This Image Bank supplements the coursework for Core 4: The Shaping of the Modern World, an introductory course offered by the History Department of Brooklyn College. The Image Bank indexes numerous historical images spanning from the Scientific Revolution to the present. The images are divided into eight major topic indexes: Ancien Regime and Critics; Age of Revolutions; Industry and Society; Liberalism and Nationalism; Varieties of Imperialism; The World Turned Upside Down; Fascism, Depression & WWII; and, The World Since 1945. Indexes for major topics contain clustered subtopics, allowing users to locate relevant images quickly. This Image Bank presents students and educators with a valuable, visual method for understanding Modern Western History.

  11. Mercury's inner core size and core crystallization regime

    NASA Astrophysics Data System (ADS)

    Dumberry, Mathieu; Rivoldini, Attilio

    2015-04-01

    Earth-based radar observation of Mercury's rotation vector combined with gravity observation by the MESSENGER spacecraft yield a measure of Mercury's moment of inertia and the amplitude of the 88-day libration of its silicate shell. These two geodetic constraints provide information on Mercury's interior structure, including the presence of a fluid core, the radius of the core-mantle boundary and the bulk densities of the core and mantle. In this work, we show how they further provide information on the size of the solid inner core and on the crystallization regime of the fluid core. If Mercury's fluid core is a Fe-FeS alloy, the largest inner core compatible with geodetic observations is 1325 ± 250 km. The crystallization scenario that best fits the observations involves the formation of Fe-snow within the fluid core. Snow formation can be restricted to a thin layer or can occupy the whole of the fluid core depending on inner core size and initial sulfur concentration. Our results offer important constraints for dynamo models of Mercury, but also advocate for the further development of models that incorporate the various features of snow formation.

  12. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect

    Bi, G.; Liu, C.; Si, S. [Shanghai Nuclear Engineering Research and Design Inst., No. 29, Hongcao Road, Shanghai, 200233 (China)

    2012-07-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)

  13. Core Multiplication in Childhood

    PubMed Central

    McCrink, Koleen; Spelke, Elizabeth S.

    2011-01-01

    A dedicated, non-symbolic, system yielding imprecise representations of large quantities (Approximate Number System, or ANS) has been shown to support arithmetic calculations of addition and subtraction. In the present study, 5–7-year-old children without formal schooling in multiplication and division were given a task requiring a scalar transformation of large approximate numerosities, presented as arrays of objects. In different conditions, the required calculation was doubling, quadrupling, or increasing by a fractional factor (2.5). In all conditions, participants were able to represent the outcome of the transformation at above-chance levels, even on the earliest training trials. Their performance could not be explained by processes of repeated addition, and it showed the critical ratio signature of the ANS. These findings provide evidence for an untrained, intuitive process of calculating multiplicative numerical relationships, providing a further foundation for formal arithmetic instruction. PMID:20537618

  14. hp calculators HP 50g House Payment Calculations

    E-print Network

    Vetter, Frederick J.

    hp calculators HP 50g House Payment Calculations The FINANCE menu House Payments Practice solving house payment calculation problems #12;hp calculators HP 50g House Payment Calculations hp calculators - 2 - HP 50g House Payment Calculations The FINANCE menu The Finance solver is accessed from the WHITE

  15. Development of New Cladding Materials Applied for Advanced LWR Aiming at Ultra-high Burn-up and Fast Neutron Spectrum

    SciTech Connect

    Kiuchi, K.; Ogawa, H.; Ioka, I. [Japan Atomic Energy Research Institute, Department of Nuclear Energy System 2-4 Shirakata Shirane, Tokai-mura, Ibaraki-ken, 319-1195 (Japan); Kuroda, Y. [The Japan Atomic Power Company, Research and Development Department (Japan); Anegawa, T. [Tokyo Electric Power Company, Nuclear Power Engineering Department (Japan)

    2002-07-01

    The ultra-high burnup more than 100 GWd/t and fast neutron spectrum tailoring are considered to be the most promising technologies applied to the advanced MOX LWRs for minimizing the electrical cost and waste management. The development of new cladding materials with the excellent irradiation properties has been conducted to realize these needs. Comparing with UO{sub 2}, to increase the internal pressure by FP gas release is accelerated with the co-production of Xe and He in MOX fuels. New stainless steels with the excellent irradiation properties, creep strength and compatibilities to high temperature water were selected to attain the reliability. The irradiation assisted stress corrosion cracking through the past experience in LWR plants is possible to inhibit by new steel making process. The problems of tritium release and PCMI is possible to inhibit by ductile niobium alloy lining. (authors)

  16. Calculating centres of mass

    E-print Network

    Vickers, James

    Calculating centres of mass 15.2 Introduction In this section we show how the idea of integration as the limit of a sum can be used to find the centre of mass of an object such as a thin plate completing this Section you should be able to . . . calculate the position of the centre of mass

  17. Statistical Tables Calculator

    NSDL National Science Digital Library

    Dinov, Ivo

    This page, created by Ino Dinov of the University of California, Berkeley, provides distribution calculators for the binomial, normal, Student's T, Chi-square, and Fisher's F distributions. Users set the parameters and enter either the probability or the test statistic and the calculators return the missing value. This is a simple, yet effective, statistical tool for instructors and students.

  18. Heat Loss Calculation Exercise

    NSDL National Science Digital Library

    Garrison, Kirk

    This class exercise from Kirk Garrison is intended for construction students learning about home insulation and heating. The class will learn to calculate heat loss in a home by using an online home heat loss calculator. This exercise document includes student worksheets. This document may be downloaded in PDF file format.

  19. Calculators and Unary Operations.

    ERIC Educational Resources Information Center

    Weaver, J. F.

    1981-01-01

    Suggests and illustrates ways in which systematic consideration of selected unary operations can be facilitated by using electronic calculators. Emphasis is placed upon unary operations suitable for exploration and investigation at the pre-algebra level, using calculation algorithms as a basis for generating examples and non-examples to develop…

  20. Modified lattice-statics approach to dislocation calculations. II - Application

    NASA Technical Reports Server (NTRS)

    Esterling, D. M.; Moriarty, J. A.

    1978-01-01

    The atomic structure of a screw dislocation core of the 110 line type in aluminum is calculated by the modified lattice-statics method developed in the preceding paper. The method includes anharmonic as well as harmonic forces and permits relaxation of the atoms in all three dimensions. All forces used in the present calculations were derived from a first-principles interatomic pair potential obtained via pseudopotential theory. Several significant differences from the ordinary lattice statics results are noted, including the displacement field, Peierl's energy barrier, and the equilibrium core-center location.