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1

Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations  

SciTech Connect

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

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

2000-03-01

2

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

Microsoft Academic Search

A rock-like oxide (ROX) fuel – light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the

H. Akie; Y. Sugo; R. Okawa

2003-01-01

3

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

SciTech Connect

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

Ohta, Hirokazu; Ogata, Takanari [Central Research Institute of Electric Power Industry, 2-11-1, Iwado Kita. Komae-shi, Tokyo 201-8511 (Japan); Obara, T. [Tokyo Institute of Technology, 2-12-1, Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

2013-07-01

4

Fast reactor 3D core and burnup analysis using VESTA  

SciTech Connect

Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

2012-07-01

5

MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION  

SciTech Connect

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

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

2011-05-08

6

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

SciTech Connect

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

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

2013-02-01

7

Methodology for embedded transport core calculation  

NASA Astrophysics Data System (ADS)

The progress in the Nuclear Engineering field leads to developing new generations of Nuclear Power Plants (NPP) with complex rector core designs, such as cores loaded partially with mixed-oxide (MOX) fuel, high burn-up loadings, and cores with advanced designs of fuel assemblies and control rods. Such heterogeneous cores introduce challenges for the diffusion theory that has been used for several decades for calculations of the current Pressurized Water Rector (PWR) cores. To address the difficulties the diffusion approximation encounters new core calculation methodologies need to be developed by improving accuracy, while preserving efficiency of the current reactor core calculations. In this thesis, an advanced core calculation methodology is introduced, based on embedded transport calculations. Two different approaches are investigated. The first approach is based on embedded finite element (FEM), simplified P3 approximation (SP3), fuel assembly (FA) homogenization calculation within the framework of the diffusion core calculation with NEM code (Nodal Expansion Method). The second approach involves embedded FA lattice physics eigenvalue calculation based on collision probability method (CPM) again within the framework of the NEM diffusion core calculation. The second approach is superior to the first because most of the uncertainties introduced by the off-line cross-section generation are eliminated.

Ivanov, Boyan D.

8

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

E-print Network

RISØ-M-2185 CALCULATION OF HEAT RATING AND BURN-UP FOR TEST FUEL PINS IRRADIATED IN DR3 C. Bagger of fuel pins irradiated in HP1 rigs. The calculations are carried out rather detailed, especially of the data. INIS Descriptors . BURN-UP, CALORIMETRY, COMPUTER CALCULATIONS, DR-3, FISSION, FUEL ASSEMBLIES

9

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

SciTech Connect

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.

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

2007-11-01

10

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

SciTech Connect

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)

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

11

Spent fuel pool storage calculations using the ISOCRIT burnup credit tool  

SciTech Connect

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.

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

2012-01-01

12

Burn-up Analysis and Determination of Equilibrium Core Configuration for Tehran Research Reactor at 7.5 MW Power Level  

SciTech Connect

This technical report presents burn-up and in-core fuel management calculations to determine a configuration for the equilibrium core for Tehran Research Reactor (TRR) at upgraded power level of 7.5 MW. Two different equilibrium core configurations have been concluded at this stage of design analysis; one equilibrium core with neutron trap and one without neutron trap. According to the preliminary fuel management calculations for the core configuration consisting of 27 Fuel Elements at 7.5 MW rated power, considering burn-up analysis, up to two neutron trap locations could be introduced in the central parts of the equilibrium core with a cycle length equal to 12 days that would satisfy the operational conditions. For the equilibrium core without neutron trap, a core with cycle length equal to 15 days gives satisfactory results. To cross-check the results of the CITVAP diffusion calculations with a Monte Carlo code such as MCNP-4B, the number densities calculated by CITVAP for the burned-up core have been provided for MCNP-4B through an auxiliary code called WIMSND. The results obtained show good agreement between these two different schemes. (authors)

Afshar, Ebrahim; Shahidi, Alireza; Zaker, Mohammad [Reactor Research and Operation Department, Nuclear Research Center Atomic Energy Organization of Iran, North Kargar Ave., P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of)

2004-07-01

13

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

SciTech Connect

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.

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

2012-12-15

14

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

NASA Astrophysics Data System (ADS)

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.

Karpushkin, T. Yu.

2012-12-01

15

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

SciTech Connect

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

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

1996-06-01

16

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

NASA Astrophysics Data System (ADS)

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

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

2014-06-01

17

CANDLE: The New Burnup Strategy  

SciTech Connect

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

Sekimoto, Hiroshi; Ryu, Kouichi; Yoshimura, Yoshikane [Tokyo Institute of Technology (Japan)

2001-11-15

18

Proliferation resistance potential and burnup characteristics of an equilibrium core of novel natural uranium fueled nuclear research reactor  

Microsoft Academic Search

Standard reactor simulation codes WIMS-D\\/4 and CITATION were employed to analyze the proliferation resistance potential and burnup characteristics of a novel natural uranium fueled nuclear research reactor [Annals of Nuclear Energy 31(12), 1331–1356]. It was found that the proposed core, which provides twice the flux per unit core power compared to similar natural uranium fueled, light water cooled, heavy water

Mohammad Javed Khan; Aslam; Nasir Ahmad

2005-01-01

19

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

SciTech Connect

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

James W. Sterbentz

2008-05-01

20

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

SciTech Connect

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.

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

1999-02-17

21

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

NASA Astrophysics Data System (ADS)

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

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

2013-10-01

22

Propagation of nuclear data uncertainties for ELECTRA burn-up calculations  

E-print Network

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

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

2013-04-08

23

Propagation of Nuclear Data Uncertainties for ELECTRA Burn-up Calculations  

NASA Astrophysics Data System (ADS)

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

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

2014-04-01

24

IN-CORE FUEL MANAGEMENT: PWR Core Calculations Using MCRAC  

NASA Astrophysics Data System (ADS)

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

Petrovi?, B. G.

1991-01-01

25

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

SciTech Connect

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

Holly R. Trellue

1998-12-01

26

ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®  

NASA Astrophysics Data System (ADS)

ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

Damian, F.; Brun, E.

2014-06-01

27

Improvements in EBR-2 core depletion calculations  

SciTech Connect

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.

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

1991-01-01

28

Propagation of Nuclear Data Uncertainties in Deterministic Calculations: Application of Generalized Perturbation Theory and the Total Monte Carlo Method to a PWR Burnup Pin-Cell  

NASA Astrophysics Data System (ADS)

The development of tools for nuclear data uncertainty propagation in lattice calculations are presented. The Total Monte Carlo method and the Generalized Perturbation Theory method are used with the code DRAGON to allow propagation of nuclear data uncertainties in transport calculations. Both methods begin the propagation of uncertainties at the most elementary level of the transport calculation - the Evaluated Nuclear Data File. The developed tools are applied to provide estimates for response uncertainties of a PWR cell as a function of burnup.

Sabouri, P.; Bidaud, A.; Dabiran, S.; Lecarpentier, D.; Ferragut, F.

2014-04-01

29

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

Microsoft Academic Search

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)

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

1997-01-01

30

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES  

SciTech Connect

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier, initial {sup 235}U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed.

BSC

2004-12-01

31

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

SciTech Connect

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)

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

32

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor  

NASA Technical Reports Server (NTRS)

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.

Butler, C.; Albright, D.

2007-01-01

33

Modifying scoping codes to accurately calculate TMI-cores with lifetimes greater than 500 effective full-power days  

SciTech Connect

The Three Mile Island unit 1 core reloads have been designed using fast but accurate scoping codes, PSUI-LEOPARD and ADMARC. PSUI-LEOPARD has been normalized to EPRI-CPM2 results and used to calculate the two-group constants, whereas ADMARC is a modern two-dimensional, two-group diffusion theory nodal code. Problems in accuracy were encountered for cycles 8 and higher as the core lifetime was increased beyond 500 effective full-power days. This is because the heavier loaded cores in both {sup 235}U and {sup 10}B have harder neutron spectra, which produces a change in the transport effect in the baffle reflector region, and the burnable poison (BP) simulations were not accurate enough for the cores containing the increased amount of {sup 10}B required in the BP rods. In the authors study, a technique has been developed to take into account the change in the transport effect in the baffle region by modifying the fast neutron diffusion coefficient as a function of cycle length and core exposure or burnup. A more accurate BP simulation method is also developed, using integral transport theory and CPM2 data, to calculate the BP contribution to the equivalent fuel assembly (supercell) two-group constants. The net result is that the accuracy of the scoping codes is as good as that produced by CASMO/SIMULATE or CPM2/SIMULATE when comparing with measured data.

Bai, D.; Levine, S.L. (Pennsylvania State Univ., University Park (United States)); Luoma, J.; Mahgerefteh, M. (GPU Nuclear Corp., Parsippany, NJ (United States))

1992-01-01

34

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

SciTech Connect

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).

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

1997-12-01

35

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

SciTech Connect

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.

DOE

1997-04-01

36

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

NASA Astrophysics Data System (ADS)

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.

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

2010-12-01

37

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

SciTech Connect

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.

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

38

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

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.

Parish, T.A.

1995-03-02

39

Calculation of Bedding Angles Inclination from Drill Core Digital Images  

Microsoft Academic Search

In this paper, we describe a new technique for the automatic orientation of bedding in drill core from digital images. Images are planar pictures of the drill core, and we show that it is possible to determine layer orientation without rotating the core on the full circumference. More precisely, we show that angle information can be ex- tracted by applying

Thomas Quiniou; Nazha Selmaoui; Christine Laporte-magoni; Michel Allenbach

2007-01-01

40

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

SciTech Connect

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)

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

41

Advances in core loss calculations for magnetic materials  

NASA Technical Reports Server (NTRS)

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.

Triner, J. E.

1982-01-01

42

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

NASA Astrophysics Data System (ADS)

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

Sloma, Tanya Noel

43

Calculation of process parameters in the manufacture of a panel with a corrugated core  

NASA Astrophysics Data System (ADS)

A method is presented for calculating the parameters of a process for producing multilayer curved panels with a lightweight corrugated core. The method provides a way to achieve precision in panel manufacture by shaping the core in accordance with the required curvature during the forming of the core. The structure and the density of the core can be varied in accordance with the desired functional application of the panels.

Khaliulin, V. I.; Desiatov, V. E.

44

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

NASA Astrophysics Data System (ADS)

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.

Sambuu, Odmaa; Nanzad, Norov

2009-03-01

45

In-core and ex-core calculations of the VENUS simulated PWR benchmark experiment  

Microsoft Academic Search

The VENUS pressurized water reactor (PWR) engineering mockup experiment was established to simulate a beginning-of-life, generic PWR configuration at the zero-power VENUS critical facility located at CEN\\/SCK in Mol, Belgium. The reactor contains Zircaloy-clad, low-enriched UOâ pins in a light water moderator. The size and pitch of the pins are consistent with modern PWR cores. The VENUS rectangular core region

M. L. Williams; P. Chowdhury; M. Landesman; F. B. K. Kam

1985-01-01

46

Local Burn-Up Effects in the NBSR Fuel Element  

SciTech Connect

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.

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

2013-01-31

47

Whole-Core Heterogeneous Transport Calculations and Their Comparison with Diffusion Results  

Microsoft Academic Search

Recently the method of characteristics (MOC) has been considered as an effective methodology in lattice calculations. This method gives accurate solutions in complex geometries and strong absorber problems. With increasingly more heterogeneous reactor cores such as a mixed-oxide (MOX) fuel-loaded core or a burnable absorber-loaded core, the limitations due to homogenization and diffusion theory are evident, and the need for

Nam Zin Cho; Gil Soo Lee; Ser Gi Hong; Chang Keun Jo; Kyung Taek Lee

2000-01-01

48

RMC - A Monte Carlo Code for Reactor Core Analysis  

NASA Astrophysics Data System (ADS)

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.

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

2014-06-01

49

The ORR Whole-Core LEU Fuel Demonstration  

SciTech Connect

The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U{sub 3}Si{sub 2}-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged {sup 235}U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of {sup 235}U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs.

Bretscher, M.M.; Snelgrove, J.L.

1990-01-01

50

Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System  

NASA Astrophysics Data System (ADS)

The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

Karim, Julia Abdul

2008-05-01

51

Benchmark Calculations for Reflector Effect in Fast Cores by Using the Latest Evaluated Nuclear Data Libraries  

NASA Astrophysics Data System (ADS)

Benchmark calculations for reflector effects in fast cores were performed to validate the reliability of scattering data of structural materials in the major evaluated nuclear data libraries, JENDL-4.0, ENDF/B-VII.1 and JEFF-3.1.2. The criticalities of two FCA and two ZPR cores were analyzed by using a continuous energy Monte Carlo calculation code. The ratios of calculation to experimental values were compared between these cores and the sensitivity analyses were performed. From the results, the replacement reactivity from blanket to SS and Na reflector is better evaluated by JENDL-4.0 than by ENDF/B-VII.1 mainly due to the ?bar values of Na and 52Cr.

Fukushima, M.; Ishikawa, M.; Numata, K.; Jin, T.; Kugo, T.

2014-04-01

52

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

SciTech Connect

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)

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

53

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

NASA Astrophysics Data System (ADS)

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.

Hartini, Entin; Andiwijayakusuma, Dinan

2014-09-01

54

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

SciTech Connect

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.

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

55

Fuel-Cycle of 'CANDLE' Burnup with Depleted Uranium  

SciTech Connect

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

Hiroshi, Sekimoto [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo (Japan)

2006-07-01

56

Calculation of Core Structure and Core Energy of a (1/2)< 111> Screw Dislocation in BCC Transition Metals: Inclusion of d-Orbital Anisotropy  

NASA Astrophysics Data System (ADS)

A tight-binding type electronic theory is used to calculate the core structure and core energy of a (1/2)< 111> screw dislocation in bcc transition metals (?-Fe, Nb and W). This calculation takes into account the d-function anisotropy of the transition metal d-band as well as the short-range repulsive energies. It is shown that the d-band effect (d-orbital anisotropy) and the next nearest-neighbour interactions are of great importance for the calculation of the core structure and core energy of the screw dislocation.

Masuda, Kin-ichi; Sato, Akikazu

1981-02-01

57

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

SciTech Connect

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.

Ellis, RJ

2001-06-01

58

Neutronic calculations for the conversion to LEU of a research reactor core  

SciTech Connect

For a five-year transitional period the Greek Research Reactor (GRR-1) was operating with a mixed core, containing both Low Enrichment (LEU) and High Enrichment (HEU) Uranium MTR- type fuel assemblies. The neutronic study of the GRR-1 conversion to LEU has been performed using a code system comprising the core-analysis code CITATION-LDI2 and the cell-calculation modules XSDRNPM and NITAWL-II of the SCALE code. A conceptual LEU core configuration was defined and analyzed with respect to the three dimensional multi-group neutron fluxes, the power distribution, the control-rod worth and the compliance with pre-defined Operation Limiting Conditions. Perturbation calculations and reactivity feedback computations were also carried out to provide input to a subsequent thermal-hydraulic study. (author)

Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus)

2008-07-15

59

First-principles calculation of H vibrational excitations at a dislocation core of Pd  

NASA Astrophysics Data System (ADS)

Palladium is an ideal system for understanding the behavior of hydrogen in metals. In Pd, H is located both in octahedral sites and in dislocation cores, which act as nanoscale H traps and form Cottrell atmospheres. Adjacent to a dislocation core, H experiences the largest possible distortion in ?-Pd . Ab initio density-functional theory computes the potential energy for a hydrogen in an octahedral site in ?-Pd and in a trap site at the core of a partial of an edge dislocation. The Pd partial dislocation core changes the environment for H, distorting the H-Pd bonding which changes the local potential, vibrational spectra, and inelastic form factor for an isolated H atom. The decrease in excitation energy is consistent with experiments, and the calculations predict distortions to the H wave function.

Lawler, Hadley M.; Trinkle, Dallas R.

2010-11-01

60

Electronic Structure Calculations and Adaptation Scheme in Multi-core Computing Environments  

SciTech Connect

Multi-core processing environments have become the norm in the generic computing environment and are being considered for adding an extra dimension to the execution of any application. The T2 Niagara processor is a very unique environment where it consists of eight cores having a capability of running eight threads simultaneously in each of the cores. Applications like General Atomic and Molecular Electronic Structure (GAMESS), used for ab-initio molecular quantum chemistry calculations, can be good indicators of the performance of such machines and would be a guideline for both hardware designers and application programmers. In this paper we try to benchmark the GAMESS performance on a T2 Niagara processor for a couple of molecules. We also show the suitability of using a middleware based adaptation algorithm on GAMESS on such a multi-core environment.

Seshagiri, Lakshminarasimhan; Sosonkina, Masha; Zhang, Zhao

2009-05-20

61

Use of Relativistic Effective Core Potentials in the Calculation of Electron-Impact Ionization Cross Sections  

NASA Technical Reports Server (NTRS)

Based on the Binary-Encounter-Bethe (BEB) model, the advantage of using relativistic effective core potentials (RECP) in the calculation of total ionization cross sections of heavy atoms or molecules containing heavy atoms is discussed. Numerical examples for Ar, Kr, Xe, and WF6 are presented.

Huo, Winifred M.; Kim, Yong-Ki

1999-01-01

62

Accurate calculation of core-electron binding energies: Multireference perturbation treatment  

NASA Astrophysics Data System (ADS)

Multireference perturbation theory (MRPT) with multiconfigurational self-consistent field (MCSCF) reference functions is applied to the calculations of core-electron binding energies (CEBEs) of atoms and molecules. Orbital relaxations in a core-ionized state and electron correlation are both taken into account in a conventional MCSCF-MRPT procedure. In the MCSCF calculation, the target core ionized state is directly optimized as an excited state and this treatment can completely prevent a variational collapse. Multireference Møller-Plesset perturbation theory and multiconfigurational self-consistent field reference quasidegenerated perturbation theory were used to treat electron correlation. The present method quite accurately reproduced the 1s CEBEs of CH4, NH3, H2O, and FH; the average deviation from the experimental data is 0.11 eV using Ahlrichs' VTZ basis set. The C 1s and O 1s CEBEs of formic acid and acetic acid were calculated and the results are consistent with the bonding characters of the atoms in these molecules. The present procedure can also be applied to CEBEs of higher angular momentum orbitals by including spin-orbit coupling. The calculated CEBEs of Ar 2p, HCl 2p, Kr 3d, and HBr 3d are in reasonable agreement with the available experimental values. In the calculation of the 3d CEBEs, a relativistic correction significantly improves the agreements. The effect of polarization functions is also discussed.

Shirai, Soichi; Yamamoto, Satoru; Hyodo, Shi-aki

2004-10-01

63

Calculation of the graphene C 1 s core level binding energy  

NASA Astrophysics Data System (ADS)

X-ray photoelectron spectroscopy combined with first-principles modeling is a powerful tool for determining the chemical composition and electronic structure of novel materials. Of these, graphene is an especially important model system for understanding the properties of other carbon nanomaterials. Here, we calculate the carbon 1 s core level binding energy of pristine graphene using two methods based on density functional theory total energy differences: a calculation with an explicit core-hole, and an all-electron extension of the delta self-consistent field (? SCF ) method. We study systematically their convergence and computational workload, and the dependence of the energies on the chosen exchange-correlation functional. The ? SCF method is computationally more expensive, but gives consistently higher C 1 s energies. Although there is a significant functional dependence, the binding energy calculated using the PBE functional is found to be remarkably close to what has been measured for graphite.

Susi, Toma; Mowbray, Duncan J.; Ljungberg, Mathias P.; Ayala, Paola

2015-02-01

64

Density functional calculation of core-electron binding energies of transition metal carbonyl and nitrosyl complexes  

NASA Astrophysics Data System (ADS)

Our recent procedure of the unrestricted generalized transition state (uGTS) model for density functional calculations of core-electron binding energies has been applied to seven carbonyl and nitrosyl inorganic complexes: Fe(CO) 5, Ni(CO) 4, Mn(CO) 4NO, Co(CO) 3NO, Fe(CO) 2(NO) 2, Mn(NO) 3CO and Cr(NO) 4. The exchange-correlation potential is based on a combined functional of Becke's exchange (B88) and Perdew's correlation (P86). The cc-pVTZ basis set was used for the calculation of neutral molecules, while for the partial cation created in the uGTS approach we scaled the cc-pVTZ basis set using a procedure based on Clementi and Raimondi's rules for atomic screening. The average absolute deviation of the calculated core-electron binding energy from experiment is 0.28 eV.

Hu, Ching-Han; Chong, Delano P.

1996-11-01

65

Examination of temperature dependent subgroup formulations in direct whole core transport calculation for power reactors  

SciTech Connect

The traditional subgroup method which has been applied for lattice transport calculations has an inherent limitation for non-uniform temperature distributions. As a measure to incorporate temperature dependence into the subgroup formulation, the subgroup level and number density adjustment method have been proposed. In this paper, the temperature dependent subgroup formulations employed for reflecting the non-uniform temperature effects on the resonance spatial self-shielding are examined for the whole core transport calculation with the thermal feedback. For 2D pin-cell problem with non-uniform temperature profiles, the inherent limitation of conventional subgroup method is confirmed. And the improvement in terms of reactivity is observed with the proposed adjustment scheme. For the real PWR core calculation with thermal feedback in the hot-full-power condition, the noticeable correction for the fuel temperature coefficient by about 10% more negative is obtained with the correction schemes. (authors)

Jung, Y. S.; Lee, U. C.; Joo, H. G. [Dept. of Nuclear Engineering, Seoul National Univ., 599 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of)

2012-07-01

66

Designing Critical Experiments in Support of Full Burnup Credit  

SciTech Connect

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.

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

2008-01-01

67

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

SciTech Connect

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

Yamamoto, T.; Suzuki, M.; Ando, Y. [Japan Nuclear Energy Safety Organization, Toranomon Towers Office, 14F, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

2012-07-01

68

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

NASA Astrophysics Data System (ADS)

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

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

2014-02-01

69

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

SciTech Connect

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)

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

70

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

PubMed

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

Kurosawa, Masahiko

2005-01-01

71

Calculation Technique for Kinematic Characteristics of Penetration of Combined Striker with Oblong Core Part Considering Possible Destruction of the Latter  

NASA Astrophysics Data System (ADS)

The up-to-date development of the armored vehicles conditions complication of armor constructions and increased slope of shell armored plates. Combined strikers (C/S) can be used to destroy armored vehicles. We can increase total weight of the core part to increase the striker's power. However, the increase of core part diameter is limited by body dimensions. Thus, we can increase core part weight by increasing its length. Because of C/S interaction with the barriers at large deviation angles, C/S's mechanical trajectory sparks in the barrier. This results in bending stress which occurs in the core part. Because of large deviation angles, the impact of the side surface of oblong core part against the cavity edge occurs. This increases the probability of core part destruction. The calculation technique for oblong core part penetration into different types of barriers is presented. The large number of factors can be calculated using this technique. It is assumed that the core part is destroyed when the tail part impacts against the cavity in the section where specific impact energy exceeds the critical value. Impact elasticity and destruction at bending stress were selected to be destruction criteria. The following core part destruction scenarios were investigated and calculated: (i) core head part is slightly destroyed but tail part of cylindrical shape penetrates deeper; (ii) core tail part is slightly destroyed but head part penetrates deeper, mass loss is taken into account; and (iii) after the impact, the core part is splitted up into two parts, then both of them penetrate into the barrier, one part is of ogival shape, the other is of cylindrical one. This calculation technique was applied to computational program, then critical angles at which core part side surface is still in contact with cavity surface, and the angles at which core part destruction occurs were calculated. Depths of core part penetration for different destruction scenarios were calculated.

Antsiferova, E. V.; Bogdanov, V. V.; Derebenko, E. V.; Lagutina, A. V.; Khmelnikov, E. A.

2006-08-01

72

Analysis of the Ikata-3 Initial Core with the CHAPLET Heterogeneous Transport Calculation Code Based on the Method of Characteristics  

Microsoft Academic Search

Improvements in cost performance of computer hardware are extending applicability of cell-heterogeneous transport calculations that are currently applied in lattice cell or assembly geometry. In this paper, a cell-heterogeneous whole-core transport calculation by the method of characteristics (MOC) is applied to calculations for analyses of pressurized water reactor (PWR) initial cores. Calculation accuracy by the MOC has been verified within

Masahiro Tatsumi; Tatsuya Kimoto; Akio Yamamoto

2000-01-01

73

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

SciTech Connect

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.

Ellis, RJ

2001-02-02

74

Atomistic breathing shell model calculations of dislocation core configurations in ionic crystals  

Microsoft Academic Search

A study was undertaken to improve upon a previous rigid boundary atomistic calculation of the core properties of a\\/2?110?{110} edge dislocations in MgO. A flexible boundary method (Flex-II) based on linear elasticity theory and recently developed by Hoagland, Hirth and Gehlen (1976) was used to provide a more accurate determination of the boundary between the atomistic and the elasto-atomic regions.

C. H. Woo; M. P. Puls

1977-01-01

75

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

SciTech Connect

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.

S. Bays; W. Skerjanc; M. Pope

2009-05-01

76

Calculation of Core Structure and Core Energy of a (1\\/2) Screw Dislocation in BCC Transition Metals: Inclusion of d-Orbital Anisotropy  

Microsoft Academic Search

A tight-binding type electronic theory is used to calculate the core structure and core energy of a (1\\/2) screw dislocation in bcc transition metals (alpha-Fe, Nb and W). This calculation takes into account the d-function anisotropy of the transition metal d-band as well as the short-range repulsive energies. It is shown that the d-band effect (d-orbital anisotropy) and the next

Kin-ichi Masuda; Akikazu Sato

1981-01-01

77

Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR  

SciTech Connect

A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.

Hanson, A.L.; Diamond, D.

2011-09-30

78

Real-time TDDFT calculations of core-hole spectral functions  

NASA Astrophysics Data System (ADS)

Core-hole response is important in a variety of x-ray spectra, including x-ray absorption, resonant and non-resonant inelastic x-ray scattering, and x-ray photo-electron spectroscopy, but has usually been treated within the adiabatic approximation. Here we explore the dynamic response of valence electrons to the sudden appearance of a deep core-hole using real time time dependent density functional theory (RT-TDDFT). The core-hole is treated as a transient time dependent potential which excites the valence electrons, as in the edge-singularity theory of Nozieres and De Dominicis. RT-TDDFT provides an efficient approach for treating response to time-dependent external fields including interactions among the valence electrons, which has recently been applied to calculations of optical and x-ray spectra.footnotetextA. J. Lee, F. D. Vila, and J. J. Rehr, PRB 86 115107 Here we generalize this approach to explore the role of the strength and localization of the core-hole potential and its effects on the spectral function and various x-ray spectra, together with comparisons to the adiabatic approximation.

Kas, J. J.; Rehr, J. J.; Lee, A. J.; Vila, F. D.

2013-03-01

79

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

E-print Network

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.

M. A. Caprio; P. Maris; J. P. Vary; R. Smith

2015-02-04

80

High accuracy modeling for advanced nuclear reactor core designs using Monte Carlo based coupled calculations  

NASA Astrophysics Data System (ADS)

The main objective of this PhD research is to develop a high accuracy modeling tool using a Monte Carlo based coupled system. The presented research comprises the development of models to include the thermal-hydraulic feedback to the Monte Carlo method and speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Presently, deterministic codes based on the diffusion approximation of the Boltzmann transport equation, coupled with channel-based (or sub-channel based) thermal-hydraulic codes, carry out the three-dimensional (3-D) reactor core calculations of the Light Water Reactors (LWRs). These deterministic codes utilize nuclear homogenized data (normally over large spatial zones, consisting of fuel assembly or parts of fuel assembly, and in the best case, over small spatial zones, consisting of pin cell), which is functionalized in terms of thermal-hydraulic feedback parameters (in the form of off-line pre-generated cross-section libraries). High accuracy modeling is required for advanced nuclear reactor core designs that present increased geometry complexity and material heterogeneity. Such high-fidelity methods take advantage of the recent progress in computation technology and coupled neutron transport solutions with thermal-hydraulic feedback models on pin or even on sub-pin level (in terms of spatial scale). The continuous energy Monte Carlo method is well suited for solving such core environments with the detailed representation of the complicated 3-D problem. The major advantages of the Monte Carlo method over the deterministic methods are the continuous energy treatment and the exact 3-D geometry modeling. However, the Monte Carlo method involves vast computational time. The interest in Monte Carlo methods has increased thanks to the improvements of the capabilities of high performance computers. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal-hydraulic models. The development of such reference high-fidelity coupled multi-physics scheme is described in this dissertation on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This work presents results from studies performed and implemented at the Pennsylvania State University (PSU) on both accelerating Monte Carlo criticality calculations by using hybrid nodal diffusion Monte Carlo schemes and thermal-hydraulic feedback modeling in Monte Carlo core calculations. The hybrid MCNP5/CTF/NEM/NJOY coupled code system is proposed and developed in this dissertation work. The hybrid coupled code system contains a special interface developed to update the required MCNP5 input changes to account for dimension and density changes provided by the thermal-hydraulics feedback module. The interface has also been developed to extract the flux and reaction rates calculated by MCNP5 to later transform the data into the power feedback needed by CTF (axial and radial peaking factors). The interface is contained in a master program that controls the flow of the calculations. Both feedback modules (thermal-hydraulic and power subroutines) use a common internal interface to further accelerate the data exchange. One of the most important steps to correctly include the thermal hydraulic feedback into MCNP5 calculations begins with temperature dependent cross section libraries. If the cross sections used for the calculations are not at the correct temperature, the temperature feedback cannot be included into MCNP5 (referred to the effect of temperature on cross sections: Doppler boarding of resolve and unresolved resonances, thermal scattering and elastic scattering). The only method of considering the temperature effects on cross sections is through the generation (or as introduced in this dissertation through a novel interpolation mechanism) of continuous energy temperature-dependent cross section libraries. An automated methodology for generation of continuous energy temperature-dependent cross section libraries has been developed

Espel, Federico Puente

81

Energy-Deposition and Damage Calculations in Core-Vessel Inserts at the Spallation Neutron Source  

SciTech Connect

Heat-deposition and damage calculations are described for core-vessel inserts in the target area of the Spallation Neutron Source. Two separate designs for these inserts (or neutron beam tubes) were studied; a single-unit insert and a multi-unit insert. The single unit contains a neutron guide; the multi unit does not. Both units are constructed of stainless steel. For the single unit, separate studies were carried out with the guide composed of stainless steel, glass, and aluminum. Results are also reported for an aluminum window on the front of the insert, a layer of nickel on the guide, a cadmium shield surrounding the guide, and a stainless steel plug in the beam-tube opening. The locations of both inserts were the most forward positions to be occupied by each design respectively thus ensuring that the calculations are conservative.

Murphy, B.D.

2002-06-25

82

First-principle calculation of core level binding energies of LixPOyNz solid electrolyte  

NASA Astrophysics Data System (ADS)

We present first-principle calculations of core-level binding energies for the study of insulating, bulk phase, compounds, based on the Slater-Janak transition state model. Those calculations were performed in order to find a reliable model of the amorphous LixPOyNz solid electrolyte which is able to reproduce its electronic properties gathered from X-ray photoemission spectroscopy (XPS) experiments. As a starting point, Li2PO2N models were investigated. These models, proposed by Du et al. on the basis of thermodynamics and vibrational properties, were the first structural models of LixPOyNz. Thanks to chemical and structural modifications applied to Li2PO2N structures, which allow to demonstrate the relevance of our computational approach, we raise an issue concerning the possibility of encountering a non-bridging kind of nitrogen atoms (=N-) in LixPOyNz compounds.

Guille, Émilie; Vallverdu, Germain; Baraille, Isabelle

2014-12-01

83

First-principle calculation of core level binding energies of LixPOyNz solid electrolyte.  

PubMed

We present first-principle calculations of core-level binding energies for the study of insulating, bulk phase, compounds, based on the Slater-Janak transition state model. Those calculations were performed in order to find a reliable model of the amorphous LixPOyNz solid electrolyte which is able to reproduce its electronic properties gathered from X-ray photoemission spectroscopy (XPS) experiments. As a starting point, Li2PO2N models were investigated. These models, proposed by Du et al. on the basis of thermodynamics and vibrational properties, were the first structural models of LixPOyNz. Thanks to chemical and structural modifications applied to Li2PO2N structures, which allow to demonstrate the relevance of our computational approach, we raise an issue concerning the possibility of encountering a non-bridging kind of nitrogen atoms (=N(-)) in LixPOyNz compounds. PMID:25554171

Guille, Émilie; Vallverdu, Germain; Baraille, Isabelle

2014-12-28

84

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

SciTech Connect

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.

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

85

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

NASA Astrophysics Data System (ADS)

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 thereactor 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 runthe 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 typeas 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.

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

2014-09-01

86

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

NASA Technical Reports Server (NTRS)

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.

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

1995-01-01

87

Thermal-hydraulic calculations for the conversion to LEU of a research reactor core  

SciTech Connect

The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Conditions. The hottest channel analysis approach was adopted, and peaking factors were used to account for fabrication or measuring uncertainties. Calculations were carried out using the numerical codes NATCON, PLTEMP and PARET provided by Argonne National Laboratory (ANL). Two main different classes of conditions were considered, namely i) steady state normal operating conditions and ii) transient cases related to accidental events including reactivity feedback effects. For steady state operating conditions the behaviour of the new configuration was examined both for forced and natural convection cooling modes. Transient calculations considered several initiating events including reactivity insertion accidents (slow or fast reactivity insertion) and total or partial loss-of-flow accidents, i.e. in accordance to guidelines provided by the IAEA for research Reactors. (author)

Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus); Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece)

2008-07-15

88

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

SciTech Connect

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

Winston, Philip Lon; Sterbentz, James William

2001-04-01

89

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

SciTech Connect

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.

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

2014-12-15

90

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

NASA Astrophysics Data System (ADS)

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.

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

2014-06-01

91

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

SciTech Connect

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)

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

92

'Dirac Fock + core-polarization' calculations of E1 transitions in the francium isoelectronic sequence  

Microsoft Academic Search

Some E1 transitions in the francium isoelectronic sequence are computed in the 'Dirac-Fock + core-polarization' approximation, where core-valence electron correlation is treated in a semiclassical core-polarization picture. The obtained ionization energies and oscillator strengths are tested versus very accurate many-body perturbation treatment (MBPT) theoretical results published recently as well as versus available experimental data. The role of core-valence correlation (core

Jacek Migdalek; Agnieszka Glowacz-Proszkiewicz

2007-01-01

93

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

SciTech Connect

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

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

2010-01-01

94

Calculations of ADS with deep subcritical uranium active cores - comparison with experiments and predictions  

NASA Astrophysics Data System (ADS)

The main characteristics of the neutron field formed within the massive (512 kg) natural uranium target assembly (TA) QUINTA irradiated by deuteron beam of JINR Nuclotron with energies 1,2,4, and 8 GeV as well as the spatial distributions and the integral numbers of (n,f), (n,?) and (n,xn)- reactions were calculated and compared with experimental data [1] . The MCNPX 27e code with ISABEL/ABLA/FLUKA and INCL4/ABLA models of intra-nuclear cascade (INC) and experimental cross-sections of the corresponding reactions were used. Special attention was paid to the elucidation of the role of charged particles (protons and pions) in the fission of natural uranium of TA QUINTA. Extensive calculations have been done for quasi-infinite (with very small neutron leakage) depleted uranium TA BURAN having mass about 20 t which are intended to be used in experiments at Nuclotron in 2014-2016. As in the case of TA QUINTA which really models the central zone of TA BURAN the total numbers of fissions, produced 239Pu nuclei and total neutron multiplicities are predicted to be proportional to proton or deuteron energy up to 12 GeV. But obtained values of beam power gain are practically constant in studied incident energy range and are approximately four. These values are in contradiction with the experimental result [2] obtained for the depleted uranium core weighting three tons at incident proton energy 0.66 GeV.

Zhivkov, P.; Furman, W.; Stoyanov, Ch

2014-09-01

95

Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses  

Microsoft Academic Search

This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that

2002-01-01

96

SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT  

SciTech Connect

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).

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

2009-01-01

97

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

NASA Astrophysics Data System (ADS)

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

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

2012-08-01

98

Model calculation of core level XPS spectra in early 3d-metal compounds  

NASA Astrophysics Data System (ADS)

Using a configuration-interaction impurity-Anderson model we calculate core-hole X-ray Photoemission Spectra (c-XPS) for some early Transition Metal (TM) insulating compounds. Because in these compounds the valence (ligand) band is completely filled, the TM on-site Coulomb repulsion U_dd is treated exactly, as well as the 3d-core hole interaction U_dc. The ground state of such a ligand-TM system with a nominally d^0 cation is described as a mixture of purely d^0 ionic state, and charge-transfer screened d^1L and d^2L^2 states where L denotes a hole in the ligand band. Our simplified model enables us to understand why c-XPS satellites are still present in CaF2 or CaO, like in light TM compounds, but absent for KF compounds. In addition to U_dd and U_dc, the other relevant parameters are the ligand-to-metal charge-transfer energy ? and the corresponding hybridization V (related to the metal-ligand transfer integrals). Finally quite a good fit to 2p{3/2}-XPS of TiO2 is obtained by using the parameter values estimated from (i) a LMTO band structure calculation of TiO2, and (ii) another calculated fit of the K(Ti) pre-edge absorption spectrum in TiO2. A l'aide du modèle d'impureté d'Anderson avec mélange de configurations, nous calculons les spectres X de la photoémission de cœur (c-XPS) pour certains composés isolants du début de la série des métaux de transition (MT). Comme la bande de valence de ces composés est complètement remplie, on peut traiter exactement la répulsion de Coulomb U_dd sur le site du MT ainsi d'ailleurs que l'interaction U_dc entre l'électron 3d et le trou de cœur. L'état de base d'un tel système contenant un cation de configuration d^0 est décrit par un mélange de l'état ionique purement d^0 et des états écrantés à transfert de charge d^1L et d^2L^2 où L est mis pour un trou dans la bande de valence. Ce modèle très simplifié est par exemple capable de comprendre pourquoi les satellites de la photoémission de cœur sont bien présents dans CaF2 ou CaO, comme dans les composés de MT légers, mais absents dans les composés KF. En plus de U_dd et U_dc, les autres paramètres pertinents sont ?, l'énergie de transfert de charge du métalloïde vers le métal et l'hybridation correspondante V (reliée aux intégrales de transferts correspondantes). Aussi, nous avons pu obtenir un bon ajustement théorique du spectre 2p{3/2}-XPS, mesuré dans TiO2, et ceci en utilisant les valeurs des paramètres fournies par (i) un calcul de structure de bandes LMTO de TiO2, (ii) un calcul des prépics du seuil K(Ti) d'absorption dans TiO2.

Parlebas, J. C.

1992-07-01

99

Technical Development on Burn-up Credit for Spent LWR Fuel  

SciTech Connect

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

Gauld, I.C.

2001-12-26

100

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet  

NASA Astrophysics Data System (ADS)

The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

2011-07-01

101

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

SciTech Connect

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.

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

102

Burnup analysis of rock-like oxide fuel disks irradiated in the Japan Research Reactor No. 3  

NASA Astrophysics Data System (ADS)

Burnup analysis of rock-like oxide (ROX) fuel disks has been carried out and the results have been compared with measured values. Two kinds of ROX disks: zirconia and thoria, were fabricated and irradiated in an irradiation hole of the Japan Research Reactor No. 3 (JRR-3). After irradiation, several post-irradiation examinations (PIE) were performed. Computer codes used for the calculations were the SRAC and the MVP-BURN codes. Firstly, the neutron spectrum in the irradiation hole was calculated using the SRAC code system. Fixed source problems were solved to obtain the neutron spectra and effective cross-sections of the disks and burnup calculations were performed. The calculated results of burnup, isotopic abundance of plutonium and production of americium and curium were compared with measurement values. Calculations overestimate the measured burnup by 7 ˜ 15% and both codes largely underestimate the measured production of americium and curium isotopes. The calculated plutonium abundance agrees moderately well with the measured values.

Nakano, Y.; Akie, H.; Magara, M.; Takano, H.

1999-08-01

103

Electronic stopping power from first-principles calculations with account for core electron excitations and projectile ionization  

NASA Astrophysics Data System (ADS)

We use Ehrenfest dynamics and time-dependent density functional theory to calculate electronic stopping power Se of energetic ions in graphitic targets from first principles. By treating core electrons as valence electrons within the projected augmented wave framework, we demonstrate that this approach provides an accurate description of Se for a wide range of ions and ion energies, even when not only valence, but also core electron excitations are essential. Our impact-parameter-dependent approach capable of describing the stopping of both low- and high-energy ions is a significant step forward in Se calculations, as it makes it possible to monitor projectile charge state during impacts, estimate contributions of core and valence electron excitations to Se, and it gives a quantitative description of electronic stopping in the cross-over region for bulk solids and nanostructures from first principles.

Ojanperä, Ari; Krasheninnikov, Arkady V.; Puska, Martti

2014-01-01

104

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

E-print Network

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 capture rates on $^{24}$Mg on an extensive temperature-density scale is presented here. This type of scale is more appropriate for interpolation purposes and of greater utility for simulation codes. The calculations are done using the pn-QRPA theory using a separable interaction. The deformation parameter, believed to be a key parameter in QRPA calculations, is adopted from experimental data to further increase the reliability of the QRPA results. The resulting calculated rates are up to a factor of 14 or more enhanced as compared to shell model rates and may lead to some interesting scenario for core collapse simulators.

Jameel-Un Nabi

2014-08-15

105

A high burnup model developed for the DIONISIO code  

NASA Astrophysics Data System (ADS)

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

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

2013-02-01

106

ATR WG-MOX Fuel Pellet Burnup Measurement by Monte Carlo - Mass Spectrometric Method  

SciTech Connect

This paper presents a new method for calculating the burnup of nuclear reactor fuel, the MCWO-MS method, and describes its application to an experiment currently in progress to assess the suitability for use in light-water reactors of Mixed-OXide (MOX) fuel that contains plutonium derived from excess nuclear weapons material. To demonstrate that the available experience base with Reactor-Grade Mixed uranium-plutonium OXide (RGMOX) can be applied to Weapons-Grade (WG)-MOX in light water reactors, and to support potential licensing of MOX fuel made from weapons-grade plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory. Fuel burnup is an important parameter needed for fuel performance evaluation. For the irradiated MOX fuel’s Post-Irradiation Examination, the 148Nd method is used to measure the burnup. The fission product 148Nd is an ideal burnup indicator, when appropriate correction factors are applied. In the ATR test environment, the spectrum-dependent and burnup-dependent correction factors (see Section 5 for detailed discussion) can be substantial in high fuel burnup. The validated Monte Carlo depletion tool (MCWO) used in this study can provide a burnup-dependent correction factor for the reactor parameters, such as capture-to-fission ratios, isotopic concentrations and compositions, fission power, and spectrum in a straightforward fashion. Furthermore, the correlation curve generated by MCWO can be coupled with the 239Pu/Pu ratio measured by a Mass Spectrometer (in the new MCWO-MS method) to obtain a best-estimate MOX fuel burnup. A Monte Carlo - MCWO method can eliminate the generation of few-group cross sections. The MCWO depletion tool can analyze the detailed spatial and spectral self-shielding effects in UO2, WG-MOX, and reactor-grade mixed oxide (RG-MOX) fuel pins. The MCWO-MS tool only needs the MS-measured 239Pu/Pu ratio, without the measured isotope 148Nd concentration data, to determine the burnup accurately. MCWO-MS not only provided linear heat generation rate, Pu isotopic composition versus burnup, and burnup distributions within the WG-MOX fuel capsules, but also correctly pointed out the inconsistency in the large difference in burnups obtained by the 148Nd method.

Chang, Gray Sen I

2002-10-01

107

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

SciTech Connect

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)

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

108

Structural and stable properties of ZnSe/Si core-shell nanowire heterostructures: The first principles calculation  

NASA Astrophysics Data System (ADS)

Relations between composition and structural and stable properties of cubic zinc selenide-silicon core-shell nanowires (NWs) are studied by first principles calculation. The diameter is between 1.1 and 2.7 nm, and the direction of the NWs considered is [110]. The lattice constants of the nanowires deviate from the Vegard's law positively with compressed ZnSe core. Stability of the NWs is discussed by taking binding energy into account. Pure Si NWs show an increasing trend of binding energy as the diameter increases while ZnSe NWs do not. Further analysis shows that zinc blende ZnSe NWs might be unstable under small diameters and a phase transition to wurtzite structure would occur. Our findings might give some guidance for the application of ZnSe/Si core-shell NWs in photoelectronics.

Zeng, Yijie; Zhou, Bofan; Huang, Yan; Fang, Yanbian; Lu, Aijiang; Wang, Chunrui; Wu, Binhe; Xu, Xiaofeng; Xing, Huaizhong

2013-12-01

109

Relativistic configuration-interaction calculation of energy levels of core-excited states in lithium-like ions: argon through krypton  

E-print Network

Large-scale relativistic configuration-interaction calculation of energy levels of core-excited states of lithium-like ions is presented. Quantum electrodynamic, nuclear recoil, and frequency-dependent Breit corrections are included in the calculation. The approach is consistently applied for calculating all $n=2$ core-excited states for all lithium-like ions starting from argon ($Z = 18$) and ending with krypton ($Z = 36$). The results obtained are supplemented with systematical estimations of calculation errors and omitted effects.

Yerokhin, V A

2012-01-01

110

Phenomena and Parameters Important to Burnup Credit  

Microsoft Academic Search

Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US

C. V. Parks; M. D. DeHart; J. C. Wagner

2001-01-01

111

MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis  

SciTech Connect

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

Gray S Chang

2005-04-01

112

Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback  

Microsoft Academic Search

Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during

S. Tashakor; G. Jahanfarnia; M. Hashemi-Tilehnoee

2010-01-01

113

Methodologies to assess potential lifetime limits for extended burnup nuclear fuel  

E-print Network

High burnup fission gas release vs. number of rods . 29 6 High burnup fission gas release vs. burnup. . 29 7 High burnup fission gas release vs. rod power. . . . . . 30 8 High burnup fission gas release vs. rod design. . . . . 30 9 High burnup... fission gas release vs. initial rod pressure. . 31 10 High burnup fission gas release vs. pellet theoretical density . 31 11 High burnup fission gas release vs. enrichment pellet . 32 12 High burnup EOL internal helium gas vs. number of rods...

De Vore, Curtis Vincent

1986-01-01

114

Depletion analysis of the UMLRR reactor core using MCNP6  

NASA Astrophysics Data System (ADS)

Accurate knowledge of the neutron flux and temporal nuclide inventory in reactor physics calculations is necessary for a variety of application in nuclear engineering such as criticality safety, safeguards, and spent fuel storage. The Monte Carlo N- Particle (MCNP6) code with integrated buildup depletion code (CINDER90) provides a high-fidelity tool that can be used to perform 3D, full core simulation to evaluate fissile material utilization, and nuclide inventory calculations as a function of burnup. The University of Massachusetts Lowell Research Reactor (UMLRR) reactor has been modeled with the deterministic based code, VENTURE and with an older version of MCNP (MCNP5). The MIT developed MCODE (MCNP ORIGEN DEPLETION CODE) was used previously to perform some limited depletion calculations. This work chronicles the use of MCNP6, released in June 2013, to perform coupled neutronics and depletion calculation. The results are compared to previously benchmarked results. Furthermore, the code is used to determine the ratio of fission products 134Cs and 137Cs (burnup indicators), and the resultant ratio is compared to the burnup of the UMLRR.

Odera, Dim Udochukwu

115

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

SciTech Connect

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)

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

116

Multipurpose Advanced 'inherently' Safe Reactor (MARS): Core design studies  

SciTech Connect

In the year 2005, in collaboration with CEA, the University of Rome 'La Sapienza' investigated a new core model with the aim at increasing the performances of the reference one, by extending the burn-up to 60 GWD/t in the case of multi-loading strategy and investigating the characteristics and limitations of a 'once-through' option, in order to enhance the proliferation resistance. In the first part of this paper, the objectives of this study and the methods of calculation are briefly described, while in the second part the calculation results are presented. (authors)

Golfier, H. [DM2S/SERMA/LCA, CEA Saclay, 91191 Gif-sur-Yvette Cedex (France); Caterino, S. [Univ. of Rome La Sapienza, Dept. of Nuclear Engineering and Energy Conversion, Corso Vittorio Emanuele II, 244 00186 Rome (Italy); Poinot, C.; Delpech, M.; Mignot, G. [DM2S/SERMA/LCA, CEA Saclay, 91191 Gif-sur-Yvette Cedex (France); Naviglio, A.; Gandini, A. [Univ. of Rome La Sapienza, Dept. of Nuclear Engineering and Energy Conversion, Corso Vittorio Emanuele II, 244 00186 Rome (Italy)

2006-07-01

117

Intrinsic electrostatic resonances of heterostructures with negative permittivity from finite-element calculations: Application to core-shell inclusions  

NASA Astrophysics Data System (ADS)

Herein, we report finite-element calculations of the effective (relative) permittivity of composite materials consisting of inclusions and inclusion arrays with a core-shell structure embedded in a surrounding host. The material making up the core of the two-dimensional structures, or cross sections of infinite three-dimensional objects (parallel, infinitely long, and identical cylinders) where the properties and characteristics are invariant along the perpendicular cross sectional plane, is assumed to have a negative real part of the permittivity, while the coating material (annular shell) is considered to be lossless. While strictly valid only in a dc situation, our analysis can be extended to treat electric fields that oscillate with time, provided that the wavelengths and attenuation lengths associated with the fields are much larger than the microstructure dimension in order that the homogeneous (effective-medium) representation of the composite structure makes sense. While one may identify features of the electrostatic resonance (ER) which are common to core-shell structures characterized by permittivities with real parts of opposite signs, it appears that the predicted ER positions are sensitive to the shell thickness and can be tuned through varying this geometric parameter. For example, we observe that the ER is broadened and shifted as the loss and the shell thickness are increased, respectively. We also argue that such core shell may also be valuable in controlling ER characteristics via polarization in an external electric field. In addition, by considering calculations of the electric field distribution, we find that the ER results in very strong and local-field enhancements into small parts of the shell perimeter. Our findings open up possibilities for the development of hybrid structures that could exploit the ER features for a particular application.

Mejdoubi, Abdelilah; Brosseau, Christian

2007-11-01

118

Three-Dimensional Full Core Power Calculations for Pressurized Water Reactors  

SciTech Connect

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.

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

2010-01-01

119

Some Nuclear Calculations of U-235-D2O Gaseous-Core Cavity Reactors  

NASA Technical Reports Server (NTRS)

The results of a multigroup, diffusion theory study of spherical gaseous-core cavity reactors are presented in this report. The reactor cavity of gaseous U235 is enclosed by a region of hydrogen gas and is separated from an external D2O moderator-reflector by a zirconium structural shell. Some cylindrical reactors are also investigated. A parametric study of spherical reactors indicates that, for the range of variables studied, critical mass increases as: (1) Fuel region is compressed within the reactor cavity, (2) moderator thickness is decreased, (3) structural shell thickness is increased, and (4) moderator temperature is increased. A buckling analogy is used to estimate the critical mass of fully reflected cylindrical reactors from spherical results without fuel compression. For a reactor cavity of a 120-centimeter radius uniformly filled with fuel, no structural shell, a moderator temperature of 70 F, and a moderator thickness of 100 centimeters, the critical mass of a spherical reactor is 3.1 kilograms while that of a cylinder with a length-to-diameter ratio of 1.0 (L/D = 1) is approximately 3.8 kilograms and, with L/D = 2, 5.9 kilograms. For the range of variables considered for U235-D2O gaseous-core cavity reactors, the systems are characterized by 95 to 99 percent thermal absorptions, with the flux reaching a maximum in the moderator about 10 to 15 centimeters from the reactor cavity.

Ragsdale, Robert G.; Hyland, Robert E.

1961-01-01

120

ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT  

SciTech Connect

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

A.H. Wells

2004-11-17

121

Calculator.  

ERIC Educational Resources Information Center

Five activities are presented in this student workbook on using the electronic calculator. Following the directions for using the machine, problems are given on multiplying and dividing, finding percentages, calculating the area of assorted polygons, changing fractions to decimals, and finding squares and square roots. (JH)

Parma City School District, OH.

122

Adaptive Calculation of a Collapsing Molecular Cloud Core: The Jeans Condition  

NASA Astrophysics Data System (ADS)

In 1997 Truelove et al. introduced the Jeans condition to determine what level of spatial resolution is needed to avoid artificial fragmentation during protostellar collapse calculations. They first found using a Cartesian code based on an adaptive mesh refinement (AMR) technique that a Gaussian cloud model collapsed isothermally to form a singular filament rather than a binary or quadruple protostellar system as predicted by previous calculations. Recently Boss et al. in 2000 using a different hydrodynamics code with high spatial resolution reproduced the filamentary collapse solution of Truelove et al., implying that high resolution coupled with the Jeans condition is necessary to perform reliable calculations of the isothermal protostellar collapse. Here we recalculate the isothermal Gaussian cloud model of Truelove et al. and Boss et al. using a completely different code based on zooming coordinates to achieve the required high spatial resolution. We follow the collapse through 7 orders of magnitude increase in density and reproduce the filamentary solution. With the zooming coordinates, we are allowed to perform an adaptive calculation with a much lower computational cost than the AMR technique and other grid redefinition methods.

Sigalotti, Leonardo Di G.; Klapp, Jaime

123

Zero-core-contribution calculation of photodetachment cross sections and photoelectron spectra of transition-metal anions  

NASA Astrophysics Data System (ADS)

The zero-core-contribution model was used to calculate (1) the absolute total photodetachment cross sections as a function of photon energies and (2) the intensities for individual detachment channels for thirteen transition-metal anions at a fixed photon energy. A unique character of each transition-metal anion is that the photodetachment could occur involving either an ``s'' or a ``d'' outermost electron. A method for determining the relative weights of the s and the d detachment orbitals is presented. The intensities calculated for different channels are in excellent agreement with photoelectron-spectroscopy measurements, both for channels involving s-orbital detachment and d-orbital detachment. Total cross sections agree with the experimental results of Feldman et al.

Stehman, R. M.; Clodius, W. B.; Grot, S.; Woo, S. B.; Helmy, E. M.

1985-01-01

124

Dependence of transuranic content in spent fuel on fuel burnup  

E-print Network

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 ...

Reese, Drew A. (Drew Amelia)

2007-01-01

125

Insights Into the PA Neutral Atom: from AN Evaluation of PA2+ Outer-Core Correlation Energy Calculations  

NASA Astrophysics Data System (ADS)

Since the identification of f-orbital contribution to the bonding in PaO+, investigations into Pa cations have hoped to characterize as many of the electronic states possible.1 Electronic states of the Pan+ (n=0-4) ions have been investigated using multi-reference spin-orbit configuration interaction (MR-SOCI). Initial investigations using Dunning style correlation consistent double-{?} basis sets are re-examined with a larger triple-{?} basis, with the hope of supporting the order of electronic states. Calculations using Hartree-Fock and CI calculations on the neutral atom did not produce the known order of states. A case study was deemed necessary on similar electron configurations present in the low energy states of Pa2+ more specifically those generated from the 5f26d1 and 5f16d2 configurations. Comparison in the Pa2+ ion is complicated by the lack of experimental results, but the states are presumed to be similar sequence as those in the neutral atom, with the addition of two electrons in the 7s shell. In evaluating the impact of inclusion of the outer core, calculations including valence-outer core correlation were completed for the 5d, 6s, and 6p shells of the Pa2+ ion. The magnitude of these individual shell correlation calculations will allow for identification of the energy level shifts associated with even and odd configurations, better describing the energy order in both the Pa2+ ion case study and for the neutral Pa atom. Upon completion of this aspect of the Pa neutral atom study, the knowledge of the energy levels in the Pan+ (n=0-4) family of ions will be greatly expanded, and may yield a model for future studies of atomic actinide systems. Gibson {et al.} Organometallics 2007, 26, 3947-3956.

Mrozik, Michael K.; Pitzer, Russell M.; Bursten, Bruce E.

2010-06-01

126

Short-term variations in core surface flow resolved from an improved method of calculating observatory monthly means  

NASA Astrophysics Data System (ADS)

Monthly means of the magnetic field measurements taken by ground observatories are a useful data source for studying temporal changes of the core magnetic field and the underlying core flow. However, the usual way of calculating monthly means as the arithmetic mean of all days (geomagnetic quiet as well as disturbed) and all local times (day and night) may result in contributions from external (magnetospheric and ionospheric) origin in the (ordinary, omm) monthly means. Such contamination makes monthly means less favourable for core studies. We calculated revised monthly means (rmm), and their uncertainties, from observatory hourly means using robust means and after removal of external field predictions, using an improved method for characterising the magnetospheric ring current. The utility of the new method for calculating observatory monthly means is demonstrated by inverting their first differences for core surface advective flows. The flow is assumed steady over three consecutive months to ensure uniqueness; the effects of more rapid changes should be attenuated by the weakly conducting mantle. Observatory data are inverted directly for a regularised core flow, rather than deriving it from a secular variation spherical harmonic model. The main field is specified by the CHAOS-4 model. Data from up to 128 observatories between 1997 and 2013 were used to calculate 185 flow models from the omm and rmm, for each possible set of three consecutive months. The full 3x3 (non-diagonal) data covariance matrix was used, and two-norm (least squares) minimisation performed. We are able to fit the data to the target (weighted) misfit of 1, for both omm and rmm inversions, provided we incorporate the full data covariance matrix, and produce consistent, plausible flows. Fits are better for rmm flows. The flows exhibit noticeable changes over timescales of a few months. However, they follow rapid excursions in the omm that we suspect result from external field contamination; this tends to cause more erratic flow speeds rather than a change in the flow pattern. We resolve temporal changes in flows derived from the rmm associated with two geomagnetic jerks that occurred around 2003.5 and 2004.5. Throughout the interval investigated, the band of westward flow straddling the equator in the hemisphere centred on the Greenwich meridian is well developed, and flows are considerably weaker beneath the Pacific Ocean. At most times, including at the start and end of our period of interest, an anti-clockwise gyre is seen beneath the southern Indian Ocean. These are the well-established long-term features of the flow. However, the gyre disappears and re-develops twice in the mid-2000s. These changes imply quite rapid and significant changes in length-of-day (assuming such changes set up torsional oscillations), which mimics changes thought to be associated with geomagnetic jerks. The bulk westward drift speed decreases throughout the interval, with oscillations superimposed. Sharp minima in 2003, 2006, 2009 and 2011 are at times Chulliat and Maus identified secular acceleration pulses at the core surface, with particularly prominent signatures at low latitudes.

Olsen, Nils; Whaler, Kathryn A.; Finlay, Christopher C.

2014-05-01

127

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

NASA Astrophysics Data System (ADS)

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.

Cabellos, O.; Sanz, J.; Rodríguez, A.; González, E.; Embid, M.; Alvarez, F.; Reyes, S.

2005-05-01

128

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

SciTech Connect

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.

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

129

Ab initio no core calculations of light nuclei and preludes to Hamiltonian quantum field theory  

SciTech Connect

Recent advances in ab initio quantum many-body methods and growth in computer power now enable highly precise calculations of nuclear structure. The precision has attained a level sufficient to make clear statements on the nature of 3-body forces in nuclear physics. Total binding energies, spin-dependent structure effects, and electroweak properties of light nuclei play major roles in pinpointing properties of the underlying strong interaction. Eventually,we anticipate a theory bridge with immense predictive power from QCD through nuclear forces to nuclear structure and nuclear reactions. Light front Hamiltonian quantum field theory offers an attractive pathway and we outline key elements.

Vary, J.P.; Maris, P.; /Iowa State U.; Shirokov, A.M.; /Iowa State U. /SINP, Moscow; Honkanen, H.; li, J.; /Iowa State U.; Brodsky, S.J.; /SLAC; Harindranath, A.; /Saha Inst.; Teramond, G.F.de; /Costa Rica U.

2009-08-03

130

Ab initio no core calculations of light nuclei and preludes to Hamiltonian quantum field theory  

SciTech Connect

Recent advances in ab initio quantum many-body methods and growth in computer power now enable highly precise calculations of nuclear structure. The precision has attained a level sufficient to make clear statements on the nature of 3-body forces in nuclear physics. Total binding energies, spin-dependent structure effects, and electroweak properties of light nuclei play major roles in pinpointing properties of the underlying strong interaction. Eventually, we anticipate a theory bridge with immense predictive power from QCD through nuclear forces to nuclear structure and nuclear reactions. Light front Hamiltonian quantum field theory offers an attractive pathway and we outline key elements.

Vary, J. P.; Maris, P.; Honkanen, H.; Li, J. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa, 50011 (United States); Shirokov, A. M. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa, 50011 (United States); Skobeltsyn Institute of Nuclear Physics, Moscow State University, Moscow, 119991 (Russian Federation); Brodsky, S. J. [SLAC National Accelerator Laboratory, Stanford University, Menlo Park, California (United States); Harindranath, A. [Theory Group, Saha Institute of Nuclear Physics, 1/AF, Bidhannagar, Kolkata, 700064 (India); Teramond, G. F. de [Universidad de Costa Rica, San Jose (Costa Rica)

2009-12-17

131

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

NASA Astrophysics Data System (ADS)

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

Erickson, Anna; Bernstein, Adam; Bowden, Nathaniel

2014-02-01

132

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

SciTech Connect

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.

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

1981-01-01

133

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

NASA Astrophysics Data System (ADS)

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.

Touran, Nicholas W.

134

Ab initio No-core Shell Model Calculations in a SU(3)-based Coupling Scheme  

NASA Astrophysics Data System (ADS)

We use powerful algorithms of computational group theory to perform ab initio configuration-interaction calculations in a SU(3)-based symmetry-adapted many-particle basis. We demonstrate that eigenfunctions for the low-lying states of 6Li, 8Be, 12C, and 16O exhibit a strong dominance of low proton, neutron, and total intrinsic spins that carry the same spatial deformation as the leading symplectic Sp(3,Bbb R) irreducible representations. Our findings imply that only a small fraction of the complete model space is needed to model nuclear collective dynamics, deformation, and ?-particle clustering even if one uses modern realistic interactions that do not preserve SU(3) symmetry. This in turn points to the importance of using a symmetry-adapted framework, one based on a LS coupling scheme with the associated spatial configurations organized according to deformation.

Dytrych, T.; Launey, K. D.; Draayer, J. P.; Langr, D.

2012-09-01

135

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

SciTech Connect

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)

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

2012-07-01

136

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

SciTech Connect

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.

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

137

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

PubMed Central

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

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

2013-01-01

138

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

PubMed

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

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

139

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

NASA Astrophysics Data System (ADS)

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.

Hirosawa, Takashi; Sato, Isamu

2011-11-01

140

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

SciTech Connect

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

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

2012-07-01

141

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

SciTech Connect

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 course 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 made clear in this report by following a sample problem.

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

1985-06-01

142

Calculation of the individual and population doses on Danish territory resulting from hypothetical core-melt accidents at the Barsebäck reactor.  

PubMed

Individual and population doses on Danish territory are calculated from hypothetical, severe core-melt accidents at the Swedish nuclear plant at Barsebäck. The release fractions for these accidents are taken from WASH-1400. Based on parametric studies, doses are calculated for very unfavourable, but not incredible weather conditions. The probability of such conditions in combination with wind direction towards Danish territory is estimated. Doses to bone marrow, lungs, GI-tract and thyroid are calculated using dose models developed at Risø. These doses are found to be consistent with doses calculated with the models used in WASH-1400. PMID:597357

Jensen, P H; Petersen, E L; Thykier-Nielsen, S; Vinther, F H

1977-10-01

143

Boundary-layer method for the analytical calculation of stable textures of bent-core liquid crystal fibers  

NASA Astrophysics Data System (ADS)

We study the equilibrium textures of molecular orientation inside cylindrical fibers made of coaxial layers of bent-core smectics. We propose a free-energy model taking into account surface-like and bulk contributions—including layer-compression and electrostatic terms among others— with constant values of the material parameters. We follow the usual variational procedure of minimization of the free energy with respect to the tilt-angle profile ?(r) and obtain an Euler-Lagrange equation and its boundary condition. We solve the variational equations for the equilibrium configurations using a boundary-layer approximation and find multiple solutions. Since the equilibrium tilt profiles are found to be radially inhomogeneous, we select those with minimum distortions in order to find the lowest free-energy state. We minimize further the free energy of the system with respect to the fiber radius and find wider intervals of stability than those previously reported, depending on the balance of the material’s spontaneous polarization, elastic and electric divergence-of-polarization constants, and surface-tension coefficients. The bulk and surface-layer structures thus found could be used to calculate the allowed modes of propagation of electromagnetic waves inside the fiber.

Pérez-Ortiz, Román; Guzmán, Orlando; Reyes, J. Adrián

2011-07-01

144

ABRAC: A microcomputer-based Fortran code for multi-cyle burnup  

SciTech Connect

Pressurized-water reactors have reactor physics and fuel management characteristics which are very amenable to simplified analysis. Given models which account for the dominant features of core and fuel performance, it is possible to rapidly perform relatively accurate scoping studies of many years of reactor operation in just a few hours on a modern (386-class) microcomputer. Models are described for burnup-dependent cross-section generation, for burnup of fuel under irradiation, and for computation of radial power distributions in hexagonal geometry assuming hexagonal fuel assemblies. Comparisons with more elaborate methods are given in order to validate the models, and to quantify the accuracy of the results. 16 refs., 5 figs., 5 tabs.

Olson, A.P.

1990-01-01

145

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

NASA Astrophysics Data System (ADS)

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.

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

2006-01-01

146

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

SciTech Connect

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.

Melissa Teague; Michael Tonks; Stephen Novascone; Steven Hayes

2014-01-01

147

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

SciTech Connect

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

Saavedra, Steven F [ORNL; Charlton, William S [Texas A& M University; Solodov, Alexander A [ORNL; Ehinger, Michael H [ORNL

2010-01-01

148

The thermonuclear burn-up in deuterated methane (CD4)  

NASA Astrophysics Data System (ADS)

The thermonuclear burn-up of highly compressed deuterated methane (CD4) is considered in the spherical geometry. The minimal required values of the burn-up parameter x = icons/Journals/Common/rho" ALT="rho" ALIGN="MIDDLE"/>0 rf are determined for various temperatures T and densities icons/Journals/Common/rho" ALT="rho" ALIGN="TOP"/>0. It is shown that thermonuclear burn-up in CD4 becomes possible in practice if its initial density icons/Journals/Common/rho" ALT="rho" ALIGN="TOP"/>0 exceeds icons/Journals/Common/approx" ALT="approx" ALIGN="TOP"/>5 × 103 g cm-3. Burn-up in CD2 T2 methane requires significantly (icons/Journals/Common/approx" ALT="approx" ALIGN="TOP"/> 100 times) lower compressions. The developed approach can be used in order to compute the critical burn-up parameters in an arbitrary fuel which contains deuterium.

Frolov, Alexei M.

1999-06-01

149

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

SciTech Connect

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

Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T. [Nuclear Energy System Safety Div., Japan Nuclear Energy Safety Organization, 4-1-28 Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

2012-07-01

150

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

SciTech Connect

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

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

1990-07-13

151

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

NASA Technical Reports Server (NTRS)

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.

Tuma, Margaret L.; Beheim, Glenn

1995-01-01

152

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

NASA Astrophysics Data System (ADS)

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

Chambon, A.; Santamarina, A.; Riffard, C.; Lavaud, F.; Lecarpentier, D.

2013-03-01

153

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

NASA Astrophysics Data System (ADS)

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

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

2014-01-01

154

On stability of spatial distributions of crystal structure defects in irradiated high burnup UO 2 fuel  

NASA Astrophysics Data System (ADS)

Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO 2 fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO 2 fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets.

Likhanskii, V. V.; Zborovskii, V. G.

2006-03-01

155

Depletion calculations for the McClellan Nuclear Radiation Center.  

SciTech Connect

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.

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

1997-12-08

156

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

SciTech Connect

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.

G. S. Chang

2005-08-01

157

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

PubMed

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

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

2008-10-01

158

Transverse buckling effects on solitary burn-up waves  

Microsoft Academic Search

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

Xue-Nong Chen; Werner Maschek

2005-01-01

159

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

Microsoft Academic Search

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

Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita; Werner Maschek

2009-01-01

160

Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations  

SciTech Connect

The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

Zhang, D.; Rahnema, F. [Georgia Institute of Technology, 770 State Street NW, Atlanta, GA 30332-0745 (United States)

2013-07-01

161

Burn-up and neutron economy of accelerator-driven reactor  

SciTech Connect

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.

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

1997-07-01

162

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

SciTech Connect

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)

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

163

Approach for Validating Actinide and Fission Product Compositions for Burnup Credit Criticality Safety Analyses  

SciTech Connect

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

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

2014-01-01

164

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

SciTech Connect

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.

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

165

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

NASA Astrophysics Data System (ADS)

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

Lemes, Martín; Soba, Alejandro; Denis, Alicia

2015-01-01

166

Generalized relativistic effective core potential and relativistic coupled cluster calculation of the spectroscopic constants for the HgH molecule and its cation  

NASA Astrophysics Data System (ADS)

Generalized relativistic effective core potential (GRECP) calculations of spectroscopic constants of the HgH molecule ground and low excited states and the HgH+ cation ground state are carried out, with correlation included by the Fock-space relativistic coupled cluster (RCC) method. Basis set superposition errors (BSSE) are estimated and discussed. It is demonstrated that connected triple excitations of the 13 outermost electrons are necessary to obtain accurate results for mercury hydride. Spectroscopic constants derived from potential curves which include these terms are in very good agreement with experiment, with errors of a few mbohr in Re, tens of wave numbers in excitation energies and vibrational frequencies, and proportionately for other properties. Comparison with previous calculations is also presented.

Mosyagin, Nikolai S.; Titov, Anatoly V.; Eliav, Ephraim; Kaldor, Uzi

2001-08-01

167

Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR using a U7Mo Dispersion Fuel  

SciTech Connect

A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The LEU fuel may be a monolithic foil (LEUm) of U10Mo (10% molybdenum by weight in an alloy with uranium) or a dispersion of U7Mo in aluminum (LEUd). A previous report provided neutronic calculations for the LEUm fuel and this report presents the neutronics parameters for the LEUd fuel. The neutronics parameters for the LEUd fuel are compared to those previously obtained for the present HEU fuel and the proposed LEUm fuel. The results show no significant differences between the LEUm and the LEUd other than the LEUd fuel requires slightly less uranium than the LEUm fuel due to less molybdenum being present. The calculations include kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions under normal operation and with misloaded fuel elements.

Hanson A. L.; Diamond D.

2014-06-30

168

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

SciTech Connect

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)

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

2013-07-01

169

Burnup increase and Power Up-rate - Operation history of KKL  

SciTech Connect

The Leibstadt nuclear power plant in Switzerland? (KKL), a GE BWR/6 boiling water reactor with an up-rated thermal power of 3600 MW and a nominal net electrical output of 1145 W has been operated for more than 20 years. The core today consists of 648 modern 10x10 assemblies with part length rods which results in a power density of 32 kW/kg Uranium or 63 kW/dm{sup 3}. The plant is operated in a 12 month cycle with shut downs in August. During the last 15 years the transformation of the core was carefully monitored and different fuel assemblies and control rods have been evaluated for an optimized performance of the core. Experience has been gained on core design with control cell core operation and a number of operational issues like change of the isothermal temperature coefficient (ITC), water chemistry with zinc addition and operation with failed fuel. The fuel and fuel assembly behavior has been monitored with numerous fuel inspections on lead use assemblies and selected reload batch assemblies. They have established a good understanding of an optimal fuel performance up to high burnup and the inspection techniques applied in the spent fuel pool on site normally during the outage. (authors)

Ledergerber, G.; Kaufmann, W.; Ritter, A.; Greiner, D. [Kernkraftwerk Leibstadt AG, CH-5325 Leibstadt (Switzerland); Parmar, Y.; Jacot-Guillarmod, R.; Krouthen, J. [Nordostschweizerische Kraftwerke AG, CH-5400 Baden (Switzerland)

2007-07-01

170

Parallelisation of MONK with Coupling to Thermal Hydraulics and Gamma Heating Calculations for Reactor Physics Applications  

NASA Astrophysics Data System (ADS)

Monte Carlo methods are increasingly being used for whole core reactor physics modelling. We describe a number of recent developments to the MONK nuclear criticality and reactor physics code to implement parallel processing, mesh-dependent burn-up and coupling to both thermal hydraulics and gamma transport codes. Results are presented which demonstrate the e_ects of gamma heating in a MONK calculation coupled to the MCBEND Monte Carlo shielding code. Experimental validation of the mesh-dependent tracking and gamma coupling methods is provided by comparison with the results of the NESSUS experiment. The gamma heating calculated by coupled MONK-MCBEND, and the neutron heating calculated by MONK, both compare well against measurement. Finally results are presented from a parallel MONK calculation of a highly detailed PWR benchmark model, which show encouraging speed-up factors on a small development cluster.

Richards, Simon D.; Davies, Nigel; Armishaw, Malcolm J.; Dobson, Geoff P.; Wright, George A.

2014-06-01

171

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

SciTech Connect

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

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

2000-03-01

172

Burnup measurements with the Los Alamos fork detector  

SciTech Connect

The fork detector system can determine the burnup of spent-fuel assemblies. It is a transportable instrument that can be mounted permanently in a spent-fuel pond near a loading area for shipping casks, or be attached to the storage pond bridge for measurements on partially raised spent-fuel assemblies. The accuracy of the predicted burnup has been demonstrated to be as good as 2% from measurements on assemblies in the United States and other countries. Instruments have also been developed at other facilities throughout the world using the same or different techniques, but with similar accuracies. 14 refs., 2 figs., 2 tabs.

Bosler, G.E.; Rinard, P.M.

1991-01-01

173

HIGH-ACCURACY MR-MP PERTURBATION THEORY ENERGY AND RADIATIVE RATES CALCULATIONS FOR CORE-EXCITED TRANSITIONS IN Fe XVI  

SciTech Connect

Accurate theoretical energy level, lifetime, and transition probability calculations of core-excited Fe XVI were performed employing the relativistic Multireference Moller-Plesset perturbation theory. In these computations the term energies of the highly excited n {<=} 5 states arising from the configuration 1s {sup 2}2s{sup k} 2p{sup m} 3l {sup p} nl' {sup q}, where k + m + p + q = 9, l {<=} 3 and p + q {<=} 2 are considered, including those of the autoionizing levels with a hole-state in the L-shell. All even and odd parity states of sodium-like iron ion were included for a total of 1784 levels. Comparison of the calculated L-shell transition wavelengths with those from laboratory measurements shows excellent agreement. Therefore, our calculation may be used to predict the wavelengths of as of yet unobserved Fe XVI, such as the second strongest 2p-3d Fe XVI line, which has not been directly observed in the laboratory and which blends with one of the prominent Fe XVII lines.

Diaz, F.; Vilkas, M. J.; Ishikawa, Y. [Department of Chemistry and the Chemical Physics Program, University of Puerto Rico, P.O. Box 23346, San Juan, PR 00931-3346 (Puerto Rico); Beiersdorfer, P., E-mail: beiersdorfer1@llnl.gov [Physics Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

2013-07-01

174

Fuel Design and Core Layout for a Gas-Cooled Fast Reactor  

SciTech Connect

The gas-cooled fast reactor (GCFR) is regarded as the primary candidate for a future sustainable nuclear power system. In this paper a general core layout is presented for a 2400-MW(thermal) GCFR. Two fuel elements are discussed: a TRISO-based coated particle and the innovative hollow sphere concept. Sustainability calls for recycling of all minor actinides (MAs) in the core and a breeding gain close to unity. A fuel cycle is designed allowing operation over a long period, requiring refueling with {sup 238}U only. The evolution of nuclides in the GCFR core is calculated using the SCALE system (one-dimensional and three-dimensional). Calculations were done over multiple irradiation cycles including reprocessing. The result is that it is possible to design a fuel and GCFR core with a breeding gain around unity, with recycling of all MAs from cycle to cycle. The burnup reactivity swing is small, improving safety. After several fuel batches an equilibrium core is reached. MA loading in the core remains limited, and the fuel temperature coefficient is always negative.

Rooijen, W.F.G. van; Kloosterman, J.L.; Hagen, T.H.J.J. van der; Dam, H. van [Delft University of Technology (Netherlands)

2005-09-15

175

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

NASA Astrophysics Data System (ADS)

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.

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

2014-06-01

176

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

SciTech Connect

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)

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

177

Application of CANDLE burnup to block-type high temperature gas cooled reactor  

Microsoft Academic Search

The CANDLE burnup strategy, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes, is applied to the block-type high temperature gas cooled reactor. If it is successful, a burnup control rod can be eliminated, and several merits are expected. This burnup may be realized

Y. Ohoka; H. Sekimoto

2004-01-01

178

About the use of the Monte-Carlo code based tracing algorithm and the volume fraction method for S n full core calculations  

SciTech Connect

The tracing algorithm that is implemented in the geometrical module of Monte-Carlo transport code MCU is applied to calculate the volume fractions of original materials by spatial cells of the mesh that overlays problem geometry. In this way the 3D combinatorial geometry presentation of the problem geometry, used by MCU code, is transformed to the user defined 2D or 3D bit-mapped ones. Next, these data are used in the volume fraction (VF) method to approximate problem geometry by introducing additional mixtures for spatial cells, where a few original materials are included. We have found that in solving realistic 2D and 3D core problems a sufficiently fast convergence of the VF method takes place if the spatial mesh is refined. Virtually, the proposed variant of implementation of the VF method seems as a suitable geometry interface between Monte-Carlo and S{sub n} transport codes. (authors)

Gurevich, M. I.; Oleynik, D. S. [RRC Kurchatov Inst., Kurchatov Sq., 1, 123182, Moscow (Russian Federation); Russkov, A. A.; Voloschenko, A. M. [Keldysh Inst. of Applied Mathematics, Miusskaya Sq., 4, 125047, Moscow (Russian Federation)

2006-07-01

179

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

SciTech Connect

The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias and uncertainty results based on a quality-assurance-controlled prerelease version of the Scale 6.1 code package and the ENDF/B-VII nuclear cross section data.

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

2011-01-01

180

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

SciTech Connect

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

Gray, W.J.

1998-06-01

181

WIMS/PANTHER analysis of UO{sub 2}/MOX cores using embedded super-cells  

SciTech Connect

This paper describes a method of analysing PWR UO{sub 2}MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

Knight, M.; Bryce, P. [EDF Energy, Barnett Way, Barnwood, Gloucester (United Kingdom); Hall, S. [Advanced Modelling and Computation Group, Imperial College, London (United Kingdom)

2012-07-01

182

Investigation of Irradiation Behavior of SiC-Coated Fuel Particle at Extended Burnup  

SciTech Connect

In current high-temperature gas-cooled reactors (HTGRs), Tri-isotropic (TRISO)-coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. The maximum burnup of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is limited to 3.6%FIMA (% fission per initial metallic atom) to certify its integrity during the operation. In order to investigate fuel behavior under extended burnup condition, irradiation tests were performed. The irradiation was carried out as HRB-22 and 91F-1A capsule irradiation tests. The fuel for the irradiation tests was called extended burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the HTTR. In order to keep fuel integrity up to over 5%FIMA, the thickness of buffer and SiC layers of fuel particle were increased. The fuel compacts were irradiated in the HRB-22 and the 91F-1A capsules at the High Flux Isotope Reactor of Oak Ridge National Laboratory and at the Japan Materials Testing Reactor of the Japan Atomic Energy Research Institute, respectively. The comparison of measured and calculated release rate-to-birth rate ratios showed that there were additional failures in both irradiation tests. A pressure vessel failure model analysis showed that no tensile stresses acted on the SiC layers even at the end of irradiation and no pressure vessel failure occurred in the intact particles even in a particle with thin buffer layer with failed OPyC layer. The presumed failure mechanisms are additional through-coatings failure of as-fabricated SiC-failed particles or an excessive increase of internal pressure by the accelerated irradiation. Further study is needed to clarify the failure mechanism.

Sawa, Kazuhiro; Tobita, Tsutomu [Japan Atomic Energy Research Institute (Japan)

2003-06-15

183

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

SciTech Connect

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).

B. Boer; A. M. Ougouag

2010-09-01

184

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 Fe-Ni alloys  

NASA Astrophysics Data System (ADS)

We have performed athermal periodic plane-wave density functional calculations within the generalised gradient approximation on the bcc, fcc and hcp structures of Fe1-XNiX alloys (X=0, 0.0625, 0.125, 0.25, and 1) in order to obtain their relative stability and elastic properties at 360 GPa and 0 K. For the hcp structure, using ab initio molecular dynamics, we have also calculated the elastic properties and wave velocities for X=0, 0.0625, and 0.125, at 360 GPa and 5500 K, with further calculations for X=0, and 0.125 at 360 GPa and 2000 K. At 0 K, the hcp structure is the most stable for X=0, 0.0625, 0.125, and 0.25, with the fcc structure becoming the most stable above X˜0.45; the bcc structure is not the most stable phase for any composition. At 0 K, compressional and shear wave velocities are structure dependent; in the case of fcc the velocities are very similar to pure Fe, but for the hcp structure the addition of Ni strongly reduces VS. Ni also reduced velocities in fcc iron, but to a lesser extent. However, at 5500 K and 360 GPa, Ni has little effect on the wave velocities of the hcp structure, which remain similar to those of pure iron throughout the range of compositions studied and, in the case of VS, >30% greater than that from seismological models. The effect of temperature on Fe-Ni alloys is, therefore, very significant, indicating that conclusions based on the extrapolation of results obtained at much lower temperatures must be treated with great caution. The significance of temperature is confirmed by the additional simulation at 2000 K for X=0, and 0.125 which reveals a remarkably linear temperature dependence of the change in VS relative to that of pure iron. At 0 K, the maximum anisotropy in VP is found to be only very weakly dependent on nickel content, but dependent on structure, being ˜15% for fcc and ˜8% for hcp. For the hcp structure at 2000 and 5500 K, the maximum anisotropy in VP is also ˜8% and almost independent of the Ni content. We conclude that Ni can safely be ignored when considering its effect on the seismic properties of hcp-Fe under core pressures and temperatures and that the negligible effect of nickel on the physical properties of iron in the core arises not because of the chemical similarities between iron and nickel, but because of the high temperature of the system.

Martorell, Benjamí; Brodholt, John; Wood, Ian G.; Vo?adlo, Lidunka

2013-03-01

185

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

SciTech Connect

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)

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

186

Calculation of fuel pin failure timing under LOCA conditions  

SciTech Connect

The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs.

Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

1991-10-01

187

Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm  

SciTech Connect

In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

Soewono, C. N.; Takaki, N. [Dept. of Applied Science Engineering, Faculty Tokai Univ., Kanagawa-ken, Hiratsuka-shi Kitakaname 4-1-1 (Japan)

2012-07-01

188

Accurate density-functional calculation of core-electron binding energies with a scaled polarized triple-zeta basis set. (III). Extension to open-shell molecules  

NASA Astrophysics Data System (ADS)

Density functional theory and the unrestricted generalized transition state (uGTS) model were applied to study the core-electron binding energies (CEBEs) of open-shell molecules. Basis set scaling based on Clementi and Raimondi's rules for atomic screening was used along with the cc-pVTZ basis set. The scaled pVTZ basis set is almost as good as the cc-pV5Z and complete basis set limit in predicting CEBEs. For small molecules (O 2, NO, NF 2 and NO 2) the average absolute deviation (aad) of our prediction (scaled pVTZ) is only 0.29 eV. For the larger molecule (CF 3) 2NO the aad is 0.56 eV, compared with experimental uncertainty of 0.5 eV. Theoretical predicted multiplet splittings for the small molecules agree quite well with experiment: the average deviation is -0.33 eV. For (CF 3) 2NO the calculated multiplet splittings are much smaller than the experimental ones. We also predict the CEBEs of PO, SN and SO, which have not been observed experimentally.

Hu, Ching-Han; Chong, Delano P.

1997-03-01

189

Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up  

NASA Astrophysics Data System (ADS)

In order to study the dependence of the gap width change on the burn-up, the fuel-to-cladding gap widths were investigated by ceramography in a large number of FBR MOX fuel pins irradiated to high burn-up. The dependence of gap widths on the burn-up was closely connected with the formations of JOG (joint oxyde-gaine) and rim structure. The gap widths decreased gradually due to the fuel swelling until ˜30 GWd/t, but beyond this burn-up the dependence showed two different tendencies. With the increase of burn-up, the gap widths decreased due to the increase of fuel swelling in the low fuel temperature region where the rim structure was observed, but they increased in the high fuel temperature region where the JOG rich in Cs and Mo formed in the gap.

Maeda, Koji; Asaga, Takeo

2004-04-01

190

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

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.

Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

2011-05-01

191

Recovery and restructuring induced by fission energy ions in high burnup nuclear fuel  

NASA Astrophysics Data System (ADS)

In light water commercial reactors, extensive change of grain structure was found at high burnup ceramic fuels. The mechanism is driven by bombardment of fission energy fragments and studies were conducted by combining accelerator based experiments and computer-science. Specimen of CeO 2 was used as simulation material of fuel ceramics. With swift heavy ion (Xe) irradiation on CeO 2, with 210 MeV, change of valence charge and lattice deviation of cations were observed by XPS and XRD. Combined irradiations of Xe implantation and swift heavy ion irradiation successfully produced sub-micrometer sized sub-grains, similar as that observed in commercial fuels. Studying components of mechanism scenarios, with first principle calculations using the VASP code, we found stable hyper-stoichiometric defect structures of UO 2+x. Molecular dynamics studies revealed stability of Xe planar defects and also found rapid transport mode of oxygen-vacancy clusters.

Kinoshita, M.; Yasunaga, K.; Sonoda, T.; Iwase, A.; Ishikawa, N.; Sataka, M.; Yasuda, K.; Matsumura, S.; Geng, H. Y.; Ichinomiya, T.; Chen, Y.; Kaneta, Y.; Iwasawa, M.; Ohnuma, T.; Nishiura, Y.; Nakamura, J.; Matzke, Hj.

2009-03-01

192

Spent LWR fuel dry storage in large transport and storage casks after extended burnup  

NASA Astrophysics Data System (ADS)

Dry spent LWR fuel storage is licensed for single fuel assemblies with rod burnup to 65 GWd/tHM. This allows dry spent fuel storage of reloads with a batch average up to 55 GWd/tHM. The leading defect mechanism for spent fuel rods in dry storage is hoop strain. Fuel rod degradation can be prevented by limiting creep. Post-pile creep of fuel rod cladding can be described conservatively by the creep of unirradiated cladding. In order to extend the database, internally pressurized creep samples were investigated for time intervals up to 10 000 h. Test temperatures were between 250 and 400°C, and the hoop stresses applied ranged from 80 to 150 N/mm 2. The resulting data were described mathematically by an interpolation formula. Based on the fuel assemblies end-of-life data the maximum CASTOR V cask storage temperature was calculated to be between 348°C and 358°C at the beginning.

Spilker, Harry; Peehs, Martin; Dyck, Hans-Peter; Kaspar, Guenter; Nissen, Klaus

1997-11-01

193

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

SciTech Connect

Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the international nuclear data community as of March 2005. The accuracy of the cross-section data was investigated by comparing existing cross-section evaluations against available measured cross-section data. When possible, benchmark calculations were also used to assess the performance of the latest FP cross-section data. Since March 2005, the U.S. and European data projects have released newer versions of their respective data files. Although there have been updates to the international data files and to some degree FP data, much of the updates have included nuclear cross-section modeling improvements at energies above the resonance region. The one exception is improved ENDF/B-VII cross-section uncertainty data or covariance data for gadolinium isotopes. In particular, ENDF/B-VII includes improved 155Gd resonance parameter covariance data, but they are based on previously measured resonance data. Although the new covariance data are available for 155Gd, the conclusions of the FP cross-section data assessment of this report still hold in lieu of the newer international cross-section data files. Based on the FP data assessment, there is judged to be a need for new total and capture cross-section measurements and corresponding cross-section evaluations, in a prioritized manner, for the nine FPs to provide the improved information and technical rigor needed for criticality safety analyses.

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

2007-12-01

194

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

E-print Network

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.

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

2005-09-15

195

Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part II: Benchmark comparisons of PUMA core parameters with MCNP5 and improvements due to a simple cell heterogeneity correction  

SciTech Connect

In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure vessel design with 451 vertical coolant channels and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update of reactor physics calculation methods and models was recently carried out covering cell, supercell (control rod) and core calculations. This paper presents benchmark comparisons of core parameters of a slightly idealized model of the Atucha-I core obtained with the PUMA reactor code with MCNP5. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, more symmetric than Atucha-II, and has some experimental data available. To validate the new models benchmark comparisons of k-effective, channel power and axial power distributions obtained with PUMA and MCNP5 have been performed. In addition, a simple cell heterogeneity correction recently introduced in PUMA is presented, which improves significantly the agreement of calculated channel powers with MCNP5. To complete the validation, the calculation of some of the critical configurations of the Atucha-I reactor measured during the experiments performed at first criticality is also presented. (authors)

Grant, C. [Comision Nacional de Energia Atomica, Av del Libertador 8250, Buenos Aires 1429 (Argentina); Mollerach, R. [Nucleoelectrica Argentina S.A., Arribenos 3619, Buenos Aires 1429 (Argentina); Leszczynski, F.; Serra, O.; Marconi, J. [Comision Nacional de Energia Atomica, Av del Libertador 8250, Buenos Aires 1429 (Argentina); Fink, J. [Nucleoelectrica Argentina S.A., Arribenos 3619, Buenos Aires 1429 (Argentina)

2006-07-01

196

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

NASA Astrophysics Data System (ADS)

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.

Su'ud, Zaki; Sekimoto, H.

2014-09-01

197

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

SciTech Connect

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.

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

198

A multi-group Monte Carlo core analysis method and its application in SCWR design  

SciTech Connect

Complex geometry and spectrum have been the characteristics of many newly developed nuclear energy systems, so the suitability and precision of the traditional deterministic codes are doubtable while being applied to simulate these systems. On the contrary, the Monte Carlo method has the inherent advantages of dealing with complex geometry and spectrum. The main disadvantage of Monte Carlo method is that it takes long time to get reliable results, so the efficiency is too low for the ordinary core designs. A new Monte Carlo core analysis scheme is developed, aimed to increase the calculation efficiency. It is finished in two steps: Firstly, the assembly level simulation is performed by continuous energy Monte Carlo method, which is suitable for any geometry and spectrum configuration, and the assembly multi-group constants are tallied at the same time; Secondly, the core level calculation is performed by multi-group Monte Carlo method, using the assembly group constants generated in the first step. Compared with the heterogeneous Monte Carlo calculations of the whole core, this two-step scheme is more efficient, and the precision is acceptable for the preliminary analysis of novel nuclear systems. Using this core analysis scheme, a SCWR core was designed based on a new SCWR assembly design. The core output is about 1,100 MWe, and a cycle length of about 550 EFPDs can be achieved with 3-batch refueling pattern. The average and maximum discharge burn-up are about 53.5 and 60.9 MWD/kgU respectively. (authors)

Zhang, P.; Wang, K.; Yu, G. [Dept. of Engineering Physics, Tsinghua Univ., Beijing, 100084 (China)

2012-07-01

199

Past experience and future needs for the use of burnup credit in LWR fuel storage  

Microsoft Academic Search

To achieve improved fuel economics and reduce the amount of fuel discharged annually, utilities are engaging in fuel management strategies that will achieve higher discharge burnups for their fuel assemblies. Although burnup credit methodologies have been developed and spent-fuel racks have been licensed, burnup credit fuel storage racks are not the answer for all utilities. Off-site and out-of-pool spent-fuel storage

W. A. Boyd; G. N. Wrights

1987-01-01

200

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

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.

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

201

Analyse de l'impact de l'environnement dans un schema de calcul a deux etapes avec DRAGON et DONJON  

NASA Astrophysics Data System (ADS)

The calculation of the neutron flux is an important data that is used to determine the dynamic of the core of a Pressurized Water Reactor (PWR). However the transport equation which gives the neutron flux, cannot be solved in three dimensions over the whole core, in evolution because of the power of the current computers, which are too slow. So some simplifications are necessary to calculate this flux. Two-levels schemes are used, where, in a first step, some macroscopic cross sections libraries are generated by solving the transport equation using infinite lattice calculations on two dimensions assemblies. These sections are generally homogenized on the whole assembly and condensed to two energy groups. In a second step, the whole core calculation is carried out using the diffusion equation, with the cross sections of the libraries previously generated, interpolated at the values of the different parameters. However the core of a PWR is made up of many assemblies, that can contain two types of fuel : Uranium OXyde (UOX) or plutonium and uranium Mixed OXyde (MOX). Moreover all these assemblies have different burnup because each one can be used for three or four cycles depending on the PWR. So that imply some burnup gradients. Thus the hypothesis of the infinite lattice used to generate the cross sections libraries can be highly inaccurate. The first goal of this project is to generate cross sections libraries that take into account the environment and to evaluate the impact of this heterogeneous environment on the core calculation. The flux obtained with the diffusion equation at the end of the core calculation is not accurate enough, du to the homogenization by assembly, to determine and to locate the hotspot factor, which represents an important industrial problematic. The principle of the power reconstruction method (PRM) is to reconstruct the more accurately possible the flux in the pins, with a combination of the diffusion flux and some microscopic flux which take into account the heterogeneities in the assemblies. This method is currently used with the data calculated with the infinite lattice. The second goal of this project is to develop a theory to apply the PRM with environmented data and to establish the PRM at the end of a calculation of the core and observe if the results are improved with the environmented data.

Bodin, Christophe

202

Monte Carlo simulation of the neutron characteristics of VVER-1000 core using the MCU-PD program and comparison of the results with calculations by the BIPR-7A program and experimental data  

SciTech Connect

The Monte Carlo method has been used to simulate the neutron transport in nuclear reactors for over fifty years. Fast progress in computer power and development of more and more robust and reliable algorithms, codes, and nuclear databases allow solving more challenging problems, including three-dimensional (3D) simulations of full-scale reactor cores. Short descriptions of a full-scale 3D model of the VVER-1000 core and algorithms and methods implemented in the MCU-PD and BIPR-7A codes and a comparison of the calculations by each program as well as a comparison with experimental data are given in this paper.

Dement'ev, V. G.; Oleinik, D. S., E-mail: oleynik@adis.vver.kiae.ru [Russian Research Centre Kurchatov Institute (Russian Federation)

2011-12-15

203

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)

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.

Triner, J. E.

1979-01-01

204

Attainable Burnup in a LIFE Engine Loaded with Depleted Uranium  

SciTech Connect

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.

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

2009-11-30

205

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

SciTech Connect

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

Enercon Services, Inc.

2011-03-14

206

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

SciTech Connect

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

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

2013-10-01

207

Impact of high-burnup fuel cycles on fuel storage systems  

Microsoft Academic Search

The utility industry's trend toward higher burnup fuel cycles (50,000 MWd\\/ton heavy metal (HM)) will change the design parameters used in the development of fuel storage systems. The overall significance of these changes (in terms of technology and economics) is not completely understood. The intent of this paper is to investigate the effects of increased initial enrichment and burnup on

J. V. Massey; B. D. Thomas; B. W. Ferguson

1986-01-01

208

U.S. Regulatory Research Program for Implementation of Burnup Credit in Transport Casks  

Microsoft Academic Search

In 1999 the U.S. Nuclear Regulatory Commission (U.S. NRC) initiated a research program to support the development of technical bases and guidance that would facilitate the implementation of burnup credit into licensing activities for transport and dry cask storage. This paper reviews the following major areas of investigation: (1) specification of axial burnup profiles, (2) assumption on cooling time, (3)

2001-01-01

209

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

Microsoft Academic Search

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

J. C. Wagner; C. V. Parks

2001-01-01

210

Effect of recrystallization in high-burnup UO{sub 2} on gas release during RIA-type transients  

SciTech Connect

The authors recently proposed a model for irradiation-induced recrystallization (grain subdivision) and swelling in UO{sub 2} fuels wherein the stored energy in the material is concentrated in a network of sink-like nuclei that diminish with dose due to interaction with radiation-produced defects. It is of considerable interest to explore the consequences of recrystallization on gas release during a reactivity initiated accident (RIA). In the absence of recrystallization, gas release during RIA-type transients is generally limited to gas available on grain boundaries and edges due to the very short heatup times (milliseconds), short cooldown periods (seconds), and relatively long intragranular diffusion distances (on the order of micrometers). However, recrystallization provides grain-boundary surfaces that are substantially closer to the gas retained in the bulk material, and thus the potential for much higher gas release. The authors show the calculated burnup at which grain subdivision will occur as a function of fractional radius and fuel temperature for a generic pressurized water reactor irradiation. The FASTGRASS code was used to calculate fission gas behavior during in-reactor irradiation and during the RIA-type transient. Results are given. It is clear from these results that recrystallization of high-burnup UO{sub 2} has implications for the potential consequences of severe accident scenarios such as the RIA type.

Rest, J.; Hofman, G.L. [Argonne National Lab., IL (United States). Energy Technology Div.

1994-10-01

211

Core-core and core-valence correlation  

NASA Technical Reports Server (NTRS)

The effect of (1s) core correlation on properties and energy separations was analyzed using full configuration-interaction (FCI) calculations. The Be 1 S - 1 P, the C 3 P - 5 S and CH+ 1 Sigma + or - 1 Pi separations, and CH+ spectroscopic constants, dipole moment and 1 Sigma + - 1 Pi transition dipole moment were studied. The results of the FCI calculations are compared to those obtained using approximate methods. In addition, the generation of atomic natural orbital (ANO) basis sets, as a method for contracting a primitive basis set for both valence and core correlation, is discussed. When both core-core and core-valence correlation are included in the calculation, no suitable truncated CI approach consistently reproduces the FCI, and contraction of the basis set is very difficult. If the (nearly constant) core-core correlation is eliminated, and only the core-valence correlation is included, CASSCF/MRCI approached reproduce the FCI results and basis set contraction is significantly easier.

Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

1988-01-01

212

VORCOR: A computer program for calculating characteristics of wings with edge vortex separation by using a vortex-filament and-core model  

NASA Technical Reports Server (NTRS)

A computer code base on an improved vortex filament/vortex core method for predicting aerodynamic characteristics of slender wings with edge vortex separations is developed. The code is applicable to camber wings, straked wings or wings with leading edge vortex flaps at subsonic speeds. The prediction of lifting pressure distribution and the computer time are improved by using a pair of concentrated vortex cores above the wing surface. The main features of this computer program are: (1) arbitrary camber shape may be defined and an option for exactly defining leading edge flap geometry is also provided; (2) the side edge vortex system is incorporated.

Pao, J. L.; Mehrotra, S. C.; Lan, C. E.

1982-01-01

213

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

NASA Astrophysics Data System (ADS)

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

Carmack, William J.

214

Development of small, fast reactor core designs using lead-based coolant.  

SciTech Connect

A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations.

Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

1999-06-11

215

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

SciTech Connect

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)

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

2012-07-01

216

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

SciTech Connect

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.

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

217

Propagation of Neutron Cross Section, Fission Yield, and Decay Data Uncertainties in Depletion Calculations  

NASA Astrophysics Data System (ADS)

Propagation of nuclear data uncertainties in reactor calculations is interesting for design purposes and libraries evaluation. Previous versions of the GRS XSUSA library propagated only neutron cross section uncertainties. We have extended XSUSA uncertainty assessment capabilities by including propagation of fission yields and decay data uncertainties due to the their relevance in depletion simulations. We apply this extended methodology to the UAM6 PWR Pin-Cell Burnup Benchmark, which involves uncertainty propagation through burnup.

Martinez, J. S.; Zwermann, W.; Gallner, L.; Puente-Espel, F.; Cabellos, O.; Velkov, K.; Hannstein, V.

2014-04-01

218

Radionuclide Data and Calculations and Loss-On-Ignition, X-Ray Fluorescence, and ICP-AES Data from Cores in Catchments of the Animas River, Colorado  

USGS Publications Warehouse

The U.S. Departments of Agriculture and Interior Abandoned Mine Lands (AML) Initiative is focused on the evaluation of the effect of past mining practices on the water quality and the riparian and aquatic habitats of impacted stream reaches downstream from historical mining districts located primarily on Federal lands. This problem is manifest in the eleven western states (west of longitude 102 degrees) where the majority of hardrock mines that had past production are located on Federal lands. In areas of temperate climate and moderate to heavy precipitation, the effects of rapid chemical and physical weathering of sulfides exposed on mine-waste dumps and acidic drainage from mines have resulted in elevated metal concentrations in the stream water and stream-bed sediment. The result of these mineral weathering processes has an unquantified impact on the quality of the water and the aquatic and riparian habitats that may limit their recreational resource value. One of the confounding factors in these studies is the determination of the component of metals derived from hydrothermally altered but unmined portions of these drainage basins. Several watersheds have been studied to evaluate the effects of acid mine drainage and acid rock drainage on the near-surface environment. The Animas River watershed in southwestern Colorado contains a large number of past-producing metal mines that have affected the watershed. Beginning in October 1996, the U.S. Geological Survey (USGS) began a collaborative study of these effects under the USGS-AML Initiative. In this report, we present the radionuclide and geochemical analytical results of sediment coring during 1997-1999 from two cores from oxbow lakes 0.5 mi. upstream from the 32nd Street Bridge near Durango, Colo., and from three cores from beaver ponds within the Mineral Creek drainage basin near Silverton, Colo.

Church, Stanley E.; Rice, Cyndi A.; Marot, Marci E.

2008-01-01

219

Design strategies for optimizing high burnup fuel in pressurized water reactors  

E-print Network

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 ...

Xu, Zhiwen, 1975-

2003-01-01

220

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

SciTech Connect

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

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

2011-01-01

221

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

E-print Network

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 ...

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

2007-01-01

222

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

SciTech Connect

One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission product LCE data to predict and verify individual biases for relevant minor actinides and fission products. This paper (1) provides a detailed description of the approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data, and (4) provides recommendations for application of the results and methods to other code and data packages.

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

2011-01-01

223

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

SciTech Connect

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

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

1982-06-01

224

Estimation of burnup in Taiwan research reactor fuel pins by using nondestructive techniques  

Microsoft Academic Search

A nondestructive measurement of spent fuel pins from the Taiwan Research Reactor has been performed at the Institute of Nuclear Energy Research. The analysis is based on a simplified balance equation for integrated flux and a series of one-group burnup-dependent microscopic cross-section libraries. A semiempirical test is used for evaluating the burnup values of two different kinds of spent fuel

Lung Kwang Pan; Cheng Si Tsao

1993-01-01

225

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

NASA Astrophysics Data System (ADS)

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

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

2007-03-01

226

Angular dependence of core hole screening in LiCoO2: A DFT+U calculation of the oxygen and cobalt K-edge x-ray absorption spectra  

E-print Network

be direclty seen from Fig. 1 at the Co K-edge in LiCoO2. HERFD-XAS is based on a two photons process (photon-electron interactions in the final state.13 However, in the case of Co K-edge in LiCoO2, it has been shown that HERFD-XASAngular dependence of core hole screening in LiCoO2: A DFT+U calculation of the oxygen and cobalt K-edge

227

Atomic force calculations within the all-electron FLAPW method: Treatment of core states and discontinuities at the muffin-tin sphere boundary  

NASA Astrophysics Data System (ADS)

We analyze the accuracy of the atomic force within the all-electron full-potential linearized augmented plane-wave (FLAPW) method using the force formalism of Yu et al. [Phys. Rev. B 43, 6411 (1991), 10.1103/PhysRevB.43.6411]. A refinement of this formalism is presented that explicitly takes into account the tail of high-lying core states leaking out of the muffin-tin sphere and considers the small discontinuities of LAPW wave function, density, and potential at the muffin-tin sphere boundaries. For MgO and EuTiO3 it is demonstrated that these amendments substantially improve the acoustic sum rule and the symmetry of the force constant matrix. Sum rule and symmetry are realized with an accuracy of ? Htr /aB .

Klüppelberg, Daniel A.; Betzinger, Markus; Blügel, Stefan

2015-01-01

228

Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY-09 Report  

SciTech Connect

The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as currently configured and used, do not yield satisfactory estimates of keff. If control rods are not modeled, these methods can deliver much better agreement with experimental core eigenvalues which suggests that development efforts should focus on modeling control rod and other absorber regions. Under some assumptions and in 1D subcore analyses, diffusion theory agrees well with transport. This suggests that developments in specific areas can produce a viable core simulation approach. Some corrections have been identified and can be further developed, specifically: treatment of the upper void region, treatment of inter-pebble streaming, and explicit (multiscale) transport modeling of TRISO fuel particles as a first step in cross section generation. Until corrections are made that yield better agreement with experiment, conclusions from core design and burnup analyses should be regarded as qualitative and not benchmark quality.

Hans D. Gougar

2009-08-01

229

Core-core and core-valence correlation  

NASA Technical Reports Server (NTRS)

The effect of 1s core correlation on properties and energy separations are analyzed using full configuration-interaction (FCI) calculations. The Be1S - 1P, the C 3P - 5S,m and CH(+) 1Sigma(+) - 1Pi separations, and CH(+) spectroscopic constants, dipole moment, and 1Sigma(+) - 1Pi transition dipole moment have been studied. The results of the FCI calculations are compared to those obtained using approximate methods.

Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

1988-01-01

230

Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core  

NASA Astrophysics Data System (ADS)

Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor 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 also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62149Sm and its dependence on the shift of a resonance position Er (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73??Er?62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant ?. We obtain new, more accurate limits of -4×10-17??·/??3×10-17yr-1. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

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

2006-12-01

231

Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core  

SciTech Connect

Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor 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 also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of {sub 62}{sup 149}Sm and its dependence on the shift of a resonance position E{sub r} (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73{<=}{delta}E{sub r}{<=}62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant {alpha}. We obtain new, more accurate limits of -4x10{sup -17}{<=}{alpha}{center_dot}/{alpha}{<=}3x10{sup -17} yr{sup -1}. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G. [St. Petersburg Nuclear Physics Institute, Gatchina, RU-188-300, St. Petersburg (Russian Federation)

2006-12-15

232

A search for minimum volume of Breed and Burn cores  

SciTech Connect

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)

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

2012-07-01

233

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan  

NASA Astrophysics Data System (ADS)

According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs). During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup. Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of an FBR. We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan. Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs. The reference specifications were selected as follows. Output of 1500MWe and average discharge fuel burnup of about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium. We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5%. The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5%. This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo. We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core.

Fujimura, Koji; Sasahira, Akira; Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi

234

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

SciTech Connect

Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in the EBR-II and study of the differences between the two fuel systems is critical for design of large advanced sodium cooled fast reactor systems. Comparing FCCI layer formation data between FFTF and EBR-II indicates that the same diffusion model can be used to represent the two systems when considering time, temperature, burnup history, and axial temperature and power profiles. This dissertation shows that FCCI formation peaks further below the top of the fuel column in FFTF experiments than has been observed in EBR-II experiments. The work provided in this dissertation will help forward the design of advanced metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full length reactor designs. This will allow the accurate lifetime prediction of fuel performance capability for new advanced sodium cooled fast reactors with extended core designs.

William J. Carmack

2012-05-01

235

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

NASA Astrophysics Data System (ADS)

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.

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

2012-01-01

236

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

SciTech Connect

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

BD Hanson; J Abrefah; SC Marschman; SG Prussin

2000-09-08

237

Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit  

SciTech Connect

Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and/or disposal and enable plants to operate with longer more economical cycle lengths and/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding material and have irradiated several test fuel rods through four cycles. The U.S. Department of Energy (DOE) joined FANP and the BWOG in supporting this project during its final phase of collecting and evaluating high burnup data through post irradiation examination (PIE).

John H. Strumpell

2004-12-31

238

Development and implementation of monitoring for the reactor core of unit No. 5 of the Novovoronezh nuclear plant by local parameters  

NASA Astrophysics Data System (ADS)

In the course of upgrading the unit no. 5 reactor core of the Novovoronezh nuclear power plant, operational limits by local parameters, which limit the admissible linear power density and the relative power of fuel elements, were established. Due to applying modern computer technologies in systems of the in-core monitoring, the calculation of power density for all fuel elements in the real-time mode is implemented. To monitor the power density of fuel elements, the algorithm for determining the limiting linear power density is developed depending on the reactor core height and on the average nuclear fuel burnup. The admissible relative power of fuel elements is determined. In the course of the performed work, the excessive conservative limitations on nonuniformity of the reactor power density are excluded. The monitoring of power density by local parameters instead of indirect K q (fuel-assembly relative power) and K v (relative power of the fuel assembly section) made it possible to increase the fuel efficiency and to improve the economic parameters of fuel cycles of the unit no. 5 reactor core of the Novovoronezh nuclear power plant.

Prytkov, A. N.; Tereshchenko, A. B.; Kravchenko, Yu. N.; Boldyrev, N. V.; Pozychaniuk, I. V.; Lisitsyn, D. I.; Golubev, E. I.

2014-04-01

239

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

SciTech Connect

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

Endo, T.; Sato, S.; Yamamoto, A. [Dept. of Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya Univ., Furo-cho, Chikusa-ku, Nagoya-shi, 464-8603 (Japan)

2012-07-01

240

Development of the core-bowing reactivity analysis code system ATLAS and its application to a large FBR core  

Microsoft Academic Search

A new code system called ATLAS has been developed to calculate core-bowing reactivity feedback behavior in fast breeder reactor cores. Based on the structural mechanics of a core (core mechanics), the code also incorporates neutron physics and thermo-hydraulic calculations. Using this new code system, the core barrel restraint system of a typical large fast breeder reactor was investigated in detail

Masatoshi Nakagawa; Hiroshi Endo; Masatoshi Kawashima

1995-01-01

241

Pore pressure and swelling in the rim region of LWR high burnup UO 2 fuel  

Microsoft Academic Search

Based on measured rim characteristics of LWR high burnup UO2 fuel, the pressure of rim pores and the additional pellet swelling due to rim formation have been modeled. Using the assumption that the number of Xe atoms retained in the rim pores is the same as that which is depleted from the rim matrix, excessive pore pressure is derived as

Yang-Hyun Koo; Byung-Ho Lee; Jin-Sik Cheon; Dong-Seong Sohn

2001-01-01

242

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

Microsoft Academic Search

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

J. Porta; J.-Y Doriath

1999-01-01

243

Estimation of burnup in Taiwan research reactor fuel pins by using nondestructive techniques  

SciTech Connect

A nondestructive measurement of spent fuel pins from the Taiwan Research Reactor has been performed at the Institute of Nuclear Energy Research. The analysis is based on a simplified balance equation for integrated flux and a series of one-group burnup-dependent microscopic cross-section libraries. A semiempirical test is used for evaluating the burnup values of two different kinds of spent fuel pins [natural uranium (0.7% [sup 235]U) and enriched uranium (7.0 % [sup 235]U)] by the [sup 134]Cs/ [sup 137]Cs activity ratio. Results are compared with radio-chemical burnup measurements. The agreement is within 3.8%, which verifies the accuracy of this method. The results are also compared with a theoretical estimation by the ORIGEN-II code. This indicates that the ORIGEN-II code's library might have an overestimated [sigma][sub a]([sup 133]Cs), which leads to a [sup 134]Cs/[sup 137]Cs ratio that would result in a burnup value [approximately]24 to 35% lower than the measured data.

Lung Kwang Pan (Chung Cheng Inst. of Tech., Taoyuan (Taiwan, Province of China)); Cheng Si Tsao (Inst. of Nuclear Energy Research, Taoyuan (Taiwan, Province of China))

1993-06-01

244

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

Microsoft Academic Search

When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with

Tanya Noel Sloma

2010-01-01

245

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

NASA Astrophysics Data System (ADS)

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 × 10 26 n/m 2 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.

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

2011-05-01

246

Object-based Interpolation via Cores  

Microsoft Academic Search

We propose an object-based interpolation that utilizes the core, a multiscale representation of object shape, as the basis for determining an interpolated object's position and intensities. The core calculations are made directly from image intensities, with no intermediate location of object boundaries. The core of the interpolated object is first determined by an interpolation of the cores of the corresponding

Derek T. Puff; David Eberly; Stephen M. Pizer?

247

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

SciTech Connect

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.

none,

2014-02-27

248

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

SciTech Connect

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

G. S. Chang

2006-07-01

249

Extended-burnup LWR (light-water reactor) fuel: The amount, characteristics, and potential effects on interim storage  

SciTech Connect

The results of a study on extended-burnup, light-water reactor (LWR) spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory for the US Department of Energy (DOE). The purpose of the study was to collect and evaluate information on the status of in-reactor performance and integrity of extended-burnup LWR fuel and initiate the investigation of the effects of extending fuel burnup on the subsequent handling, interim storage, and other operations (e.g., rod consolidation and shipping) associated with the back end of the fuel cycle. The results of this study will aid DOE and the nuclear industry in assessing the effects on waste management of extending the useful in-reactor life of nuclear fuel. The experience base with extended-burnup fuel is now substantial and projections for future use of extended-burnup fuel in domestic LWRs are positive. The basic performance and integrity of the fuel in the reactor has not been compromised by extending the burnup, and the potential limitations for further extending the burnup are not severe. 104 refs., 15 tabs.

Bailey, W.J.

1989-03-01

250

Accurate density-functional calculation of core-electron binding energies with a scaled polarized triple-zeta basis set. Twelve test cases and application to three C 2H 4O 2 isomers  

NASA Astrophysics Data System (ADS)

A scaling procedure based on Clementi and Raimondi's rules for atomic screening was proposed for atomic orbital basis sets in the unrestricted generalized transition state (uGTS) model of density functional calculation of core-electron binding energies (CEBEs). The exchange-correlation potential is based on a combined functional of Becke's exchange (B88) and Perdew's correlation (P86). This proposal was tested on CEBEs of twelve small molecules, including F 2, N 2 and H 2O, and applied to the computation of CEBEs of three isomers of C 2H 4O 2: acetic acid (CH 3COOH), methyl formate (HCOO?CH 3), and glycolic aldehyde (CH 2OHCHO). In all cases, the new scaled pVTZ basis performs almost as well as the much larger cc-pV5Z and the average absolute difference between the results from the scaled pVTZ and estimated complete basis set limits is 0.04 eV.

Chong, Delano P.; Hu, Ching-Han; Duffy, Patrick

1996-02-01

251

DUBLIN CORE  

EPA Science Inventory

The Dublin Core is a metadata element set intended to facilitate discovery of electronic resources. It was originally conceived for author-generated descriptions of Web resources, and the Dublin Core has attracted broad ranging international and interdisciplinary support. The cha...

252

Viscosity of the Earth's Core  

Microsoft Academic Search

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;

Roger F. Gans

1972-01-01

253

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

SciTech Connect

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

Kurchenkov, A. Yu., E-mail: s327@vver.kiae.ru [Russian Research Centre Kurchatov Institute (Russian Federation)

2011-12-15

254

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

Federal Register 2010, 2011, 2012, 2013, 2014

...Burnup Dry Storage Cask Research and Development Project...the Electric Power research Institute (EPRI) team, submits a License...Nuclear Fuel Disposition Research and Development has...coordinated this effort in collaboration with its...

2013-11-12

255

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

SciTech Connect

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

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

1997-12-01

256

Dislocation core-core interaction and Peierls stress in a model hexagonal lattice  

Microsoft Academic Search

A series of atomistic calculations is performed in order to explore dislocation core-core interactions and the Peierls stress in a model hexagonal lattice. The method of calculation is the lattice Green's function method, using several pair potentials with variable parameters. We confirm that dislocation cores broaden as a pair of dislocations with opposite sign move closer to each other. Continuum

S. J. Zhou; A. E. Carlsson; Robb Thomson

1994-01-01

257

Dislocation core-core interaction and Peierls stress in a model hexagonal lattice  

Microsoft Academic Search

A series of atomistic calculations is performed in order to explore dislocation core-core interactions and the Peierls stress in a model hexagonal lattice. The method of calculation is the lattice Green`s function method, using several pair potentials with variable parameters. We confirm that dislocation cores broaden as a pair of dislocations with opposite sign move closer to each other. Continuum

S. J. Zhou; A. E. Carlsson; Robb Thomson

1994-01-01

258

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

SciTech Connect

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.

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

2006-10-31

259

Starless Cores  

E-print Network

Dense low mass cores in nearby clouds like Taurus and Auriga are some of the simplest sites currently forming stars like our Sun. Because of their simplicity and proximity, dense cores offer the clearest view of the different phases of star formation, in particular the conditions prior to the onset of gravitational collapse. Thanks to the combined analysis of the emission from molecular lines and the emission/absorption from dust grains, the last several years have seen a very rapid progress in our understanding of the structure and chemical composition of starless cores. Previous contradictions between molecular tracers are now understood to arise from core chemical inhomogeneities, which are caused by the selective freeze out of molecules onto cold dust grains. The analysis of the dust emission and absorption, in addition, has allowed us to derive accurate density profiles, and has made finally possible to carry out self consistent modeling of the internal structure of starless cores. In this paper I briefly review the evolution of core studies previous to the current golden age, and show how multi-tracer emission can now be modeled in a systematic manner. Finally I show how we can start to reconstruct the early history of core formation taking advantage of the chemical changes in the gas.

Mario Tafalla

2005-04-23

260

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

SciTech Connect

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.

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

2003-09-15

261

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

NASA Astrophysics Data System (ADS)

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.

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

262

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

E-print Network

Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact of different covariance data is studied by comparing two of the presently most complete nuclear data covariance libraries (ENDF/B-VII.1 and SCALE 6.0), which reveals a high dependency of the uncertainty estimates on the source of covariance data. The burn-up benchmark Exercise I-1b proposed by the OECD expert group "Benchmarks for Uncertainty Analysis in Modeling (UAM) for the Design, Operation and Safety Analysis of LWRs" is studied as an example application. The burn-up simulations are performed with the SCALE 6.0 tool suite.

Carlos Javier Diez; Oliver Buss; Axel Hoefer; Dieter Porsch; Oscar Cabellos

2014-11-04

263

24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES ...  

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

24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES FOR A BRASS GATE VALVE BODY MADE ON A CORE BOX, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

264

McCARD for Neutronics Design and Analysis of Research Reactor Cores  

NASA Astrophysics Data System (ADS)

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.

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

2014-06-01

265

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

NASA Astrophysics Data System (ADS)

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).

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

266

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

NASA Astrophysics Data System (ADS)

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

Ramamoorthy, Karthikeyan

267

Accuracy of Monte Carlo Criticality Calculations During BR2 Operation  

SciTech Connect

The Belgian Material Test Reactor BR2 is a strongly heterogeneous high-flux engineering test reactor at SCK-CEN (Centre d'Etude de l'Energie Nucleaire) in Mol with a thermal power of 60 to 100 MW. It deploys highly enriched uranium, water-cooled concentric plate fuel elements, positioned inside a beryllium reflector with a complex hyperboloid arrangement of test holes. The objective of this paper is to validate the MCNP and ORIGEN-S three-dimensional (3-D) model for reactivity predictions of the entire BR2 core during reactor operation. We employ the Monte Carlo code MCNP-4C to evaluate the effective multiplication factor k{sub eff} and 3-D space-dependent specific power distribution. The one-dimensional code ORIGEN-S is used to calculate the isotopic fuel depletion versus burnup and to prepare a database with depleted fuel compositions. The approach taken is to evaluate the 3-D power distribution at each time step and along with the database to evaluate the 3-D isotopic fuel depletion at the next step and to deduce the corresponding shim rod positions of the reactor operation. The capabilities of both codes are fully exploited without constraints on the number of involved isotope depletion chains or an increase of the computational time. The reactor has a complex operation, with important shutdowns between cycles, and its reactivity is strongly influenced by poisons, mainly {sup 3}He and {sup 6}Li from the beryllium reflector, and the burnable absorbers {sup 149}Sm and {sup 10}B in the fresh UAl{sub x} fuel. The computational predictions for the shim rod positions at various restarts are within 0.5 $ ({beta}{sub eff} = 0.0072)

Kalcheva, Silva; Koonen, Edgar; Ponsard, Bernard [SCK-CEN (Belgium)

2005-08-15

268

Graphite disk UO/sub 2/ fuel elements designed for extended burnups at high powers  

SciTech Connect

Zircaloy-clad fuel elements containing UO/sub 2/ pellets separated by thin graphite disks have been irradiated to maximum burnups of 800 MWh/kg U (33 000 MWd/tonne U) at a linear power range of 30 to 70 kW/m. Fission gas release and sheath strains were lower than experienced for conventional fuel under comparable conditions because of the lower bulk average fuel temperatures in disk elements. The irradiated disk elements also showed good internal stability, tolerance to power ramping, and acceptable defect behavior.

MacDonald, R.D.; Hastings, I.J.

1985-11-01

269

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

PubMed

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

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

2012-06-01

270

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

SciTech Connect

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.

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

271

An Analysis of Nuclear Fuel Burnup in the AGR 1 TRISO Fuel Experiment Using Gamma Spectrometry, Mass Spectrometry, and Computational Simulation Techniques  

SciTech Connect

AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1 %FIMA for the direct method and 20.0 %FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3 % FIMA to 10.7 % FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. The results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry for TRISO fuel compacts across a burnup range of approximately 10 to 20 % FIMA and also validate the approach used in the physics simulation of the AGR 1 experiment.

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

2014-10-01

272

A Feasibility and Optimization Study to Design a Nondestructive ATR Fuel Permanent Scanning System to Determine Fuel Burnup  

NASA Astrophysics Data System (ADS)

The goal of this project was to develop the best available non-destructive technique to determine burnup of the Advanced Test Reactor (ATR) fuels at Idaho National Laboratory, as well as to make a recommendation regarding the feasibility of implementing a permanent fuel scanning system at the ATR canal. The study determined that useful spectra for validation and fuel burnup predictions can be obtained in-situ at the ATR canal using three different detectors. In addition, the study established that calibration curves can be created to predict ATR fuel burnup onsite. The study also established that in order to design a rugged system that can stand the daily operations at the ATR canal a LaBr3 scintillator can be used effectively if deconvolution process is applied to increase the spectra resolution.

Navarro, J.; Ring, T. A.; Nigg, D. W.

2014-04-01

273

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

NASA Astrophysics Data System (ADS)

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

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

2010-05-01

274

Ethics CORE  

NSDL National Science Digital Library

The Ethics CORE Digital Library, funded by the National Science Foundation, "brings together information on best practices in research, ethics instruction and responding to ethical problems that arise in research and professional life." It's a remarkable site where visitors can make their way through ethics resources for dozens of different professions and activities. The Resources by Discipline area is a great place to start. Here you will find materials related to the biological sciences, business, computer & information science, along with 14 additional disciplines. The Current News area is a great place to learn about the latest updates from the field. Of note, these pieces can easily be used in the classroom or shared with colleagues. The dynamism of the site can be found at the Interact with Ethics CORE area. Active learning exercises can be found here, along with instructional materials and visitors' own lessons learned.

275

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

NASA Astrophysics Data System (ADS)

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

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

2013-02-01

276

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

SciTech Connect

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.

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

2014-01-01

277

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

NASA Astrophysics Data System (ADS)

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.

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

2012-01-01

278

MCU-FCP program for first-collisions probability calculations of neutron transport  

Microsoft Academic Search

A general scheme, methods, and algorithms, implemented in the MCU-FCP program, making it possible to calculate the neutron-physical\\u000a characteristics of two-and three-dimensional RBMK cells and polycells taking account of fuel burnup during reactor operation\\u000a by the first-collisions probability method, are described. Test calculations are preformed and the results are compared with\\u000a the MCU-REA program.

E. A. Gomin; M. I. Gurevich; M. A. Kalugin; M. S. Yudkevich; A. P. Zhirnov; I. M. Rozhdestvenskii

2008-01-01

279

Thermal conductivity of homogeneous and heterogeneous MOX fuel with up to 44 MWd/kgHM burn-up  

NASA Astrophysics Data System (ADS)

New thermal diffusivity data for homogeneous SBR and heterogeneous MIMAS and OCOM MOX fuels are reported. No significant difference between the thermal diffusivity of the homogeneous and heterogeneous fuels was found at the burn-up up to 44 MWd/kgHM. These measurements, combined with previously published results or correlation functions for irradiated UO 2 and MOX were compared and it was found that separate correlations for these two fuels are not justified. A correlation for the thermal conductivity of irradiated UO 2 and MOX as a function of burn-up and irradiation temperature is proposed.

Staicu, D.; Cozzo, C.; Pagliosa, G.; Papaioannou, D.; Bremier, S.; Rondinella, V. V.; Walker, C. T.; Sasahara, A.

2011-05-01

280

PRESTELLAR CORES IN THE COALSACK  

SciTech Connect

We present high spectral resolution millimeter mapped observations of seven prestellar cores in the Coalsack, including imaging in five optically thin molecular species of the kinematic structure of two of the densest cores, C2 and C4. Various collapse-critical indices are calculated; critical masses needed for collapse are consistently greater than those observed, the latter ranging from 0.4 to 2.4 M{sub sun}. The molecular emission in several of the cores shows line profiles with infall characteristics as well as elongated areas of increased line widths and reversals of center velocity gradients, implying that accretion disks may be forming.

Saul, M.; Cunningham, M. [School of Physics, University of New South Wales, Sydney, 2052 NSW (Australia); Rathborne, J. [Departamento de Astronomia, Las Condes, Santiago (Chile); Walsh, W. [Harvard Smithsonian Center for Astrophysics, Cambridge, MA 02138 (United States); Butner, H. M., E-mail: msaul@phys.unsw.edu.au, E-mail: mariac@phys.unsw.edu.au, E-mail: rathborn@das.uchile.cl, E-mail: wwalsh@cfa.harvard.edu, E-mail: butnerhm@jmu.edu [Department of Physics and Astronomy, James Madison University, Harrisonburg, VA 22807 (United States)

2011-09-10

281

Gravity Calculator  

NSDL National Science Digital Library

The gravity calculator calculates the gravitational force between two masses. Also included is a visualization of the typical measurement of gravitational force (weight) in different environments (stationary and free fall).

Brendan Cannell, Ronnie Johnson, The Shodor Education Foundation, Inc.

282

Bayesian Calculator  

NSDL National Science Digital Library

This page, created by Michael H. Birnbaum of Fullerton University, uses Bayes' Theorem to calculate the probability of a hypothesis given a datum. An example about cancer is given to help users understand Bayes' Theorem and the calculator. This page is a great representation of conditional probability. Detailed instructions are provided on proper use of the calculator.

Birnbaum, Michael H.

283

Analyses of Greek Research Reactor with mixed HEU-LEU Be reflected core  

SciTech Connect

The fuel-cycle analyses presented in this paper provide specific steps to be taken in the transition from a 36-element water-reflected HEU core to a 33-element LEU equilibrium core with a Be reflector on two faces. The first step will be to install the Be reflector and remove the highest burnup HEU fuel. The smaller Be-reflected core will be refueled with LEU fuel. All analyses were performed using a planar 5-group REBUS3 model benchmarked to VIM Monte Carlo. In addition to fuel cycle results, the control rod worth, reactivity response to increased fuel and water temperature and decreased water density were compared for the transition core and the reference HEU core.

Deen, J.R.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Papastergiou, K. [National Center for Scientific Research, Athens (Greece)

1993-12-31

284

Taking burnup credit into account in criticality studies: the situation as it is now and the prospect for the future  

Microsoft Academic Search

As enrichment of the fuel has become higher than the limits used at the designing stages, it seemed necessary to consider fuel depletion during irradiation to guarantee the criticality safety for relatively high enriched fuels transportation, storage or reprocessing. This burnup credit will make it possible to use the devices for spent fuels which are initially relatively high enriched. For

P Cousinou; C Lavarenne; D Biron; M Douchet; J. P Grouiller; N Thiollay; E Guillou

2001-01-01

285

Metallic inert matrix fuel concept for minor actinides incineration to achieve ultra-high burn-up  

NASA Astrophysics Data System (ADS)

The advantages of using Inert Matrix Fuel (IMF) in a design of an isolated arrangement of fuel are considered, with emphasis on, low temperatures in the fuel center, achievement of high burn-ups, and an environment friendly process for the fuel element fabrication. Changes in the currently existing concept of IMF usage are suggested, involving novel IMF design in the nuclear fuel cycle.

Lipkina, K.; Savchenko, A.; Skupov, M.; Glushenkov, A.; Vatulin, A.; Uferov, O.; Ivanov, Y.; Kulakov, G.; Ershov, S.; Maranchak, S.; Kozlov, A.; Maynikov, E.; Konova, K.

2014-09-01

286

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

SciTech Connect

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.

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

2012-12-15

287

Automated Core Design  

SciTech Connect

Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process.

Kobayashi, Yoko; Aiyoshi, Eitaro

2005-07-15

288

Procedures for predicting pressures inside cores  

NASA Astrophysics Data System (ADS)

Core gas defects are among the most aggravating defects because they are difficult to control and may not be found until castings are machined. These defects occur when the pressure in the core is higher than the external pressure acting on the core from the metal-head pressure. The prediction of core and mold gas defects requires a determination of the permeability of the cores and the rate and volume of gas evolved from the cores in contact with molten metal. Techniques were developed for measuring core permeability and gas evolution. Gas permeability was measured at pressure levels that are seen inside cores during casting. The volume and rate of gas evolution from cores submerged in molten metal was also measured. Use of these techniques was demonstrated on commercial cores that were submerged in molten aluminum and iron. The effects of common core-making variables and casting temperatures were determined. Permeability depended mainly on compaction level, with increased density associated with reduced permeability. Coatings decreased permeability, while binder, sand type, and additives had no affect. Gas evolution volumes and rates for cores immersed in molten metal were higher in phenolic urethane cold box cores than in epoxy acrylic cores. Higher binder content, additives, coatings, immersion temperatures, core length, and metal contact area all increased evolved gas volumes and rates. A method for calculating the core pressure in simple geometries was developed and confirmed experimentally. The data generated from the gas evolution measurements were used to build a physical model on binder decomposition and the resultant gas evolution during casting. This model was used to determine the amount of gas evolved from cores at various geometries and temperatures. The model accurately predicted the volume of gas evolved. However, the composition of the gases, the core temperature profile, and more precise interfacial heat transfer and sand thermal conductivity data are also required to match the experimental rate curves.

Winardi, Leonard

289

HTTF Core Stress Analysis  

SciTech Connect

In accordance with the need to determine whether cracking of the ceramic core disks which will be constructed and used in the High Temperature Test Facility (HTTF) for heatup and cooldown experiments, a set of calculation were performed using Abaqus to investigate the thermal stresses levels and likelihood for cracking. The calculations showed that using the material properties provided for the Greencast 94F ceramic, cracking is predicted to occur. However, this modeling does not predict the size or length of the actual cracks. It is quite likely that cracks will be narrow with rough walls which would impede the flow of coolant gases entering the cracks. Based on data recorded at Oregon State University using Greencast 94F samples that were heated and cooled at prescribed rates, it was concluded that the likelihood that the cracks would be detrimental to the experimental objectives is small.

Brian D. Hawkes; Richard Schultz

2012-07-01

290

CALCULATION OF THE NEUTRON NOISE INDUCED BY SHELL-MODE  

E-print Network

CALCULATION OF THE NEUTRON NOISE INDUCED BY SHELL-MODE FISSION REACTORS CORE-BARREL VIBRATIONS IN A KEYWORDS: core-barrel vibra tions, in-core neutron noise, shell- mode vibrations 1-D, TWO-GROUP, TWO by the shell-mode vibra tions of the core barrel. The original motivation was to investigate whether an out

Demazière, Christophe

291

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.

292

Helium release from the uranium-plutonium mixed oxide (MOX) fuel irradiated to high burn-up in a fast breeder reactor (FBR)  

NASA Astrophysics Data System (ADS)

The helium releases were investigated in FBR fuel pins irradiated to high burn-up. The released amounts increased with the increase of burn-up, but no burn-up dependence of their release fraction to total amounts generated was seen. The contents of 241Am in fuel pellets affected the released amounts. The effect of released helium on the internal pressure in fuel pins was smaller in FBR fuel pins than in LWR fuel pins, but released helium in fuel pins containing increased amounts of 241Am, which are anticipated in future FBRs, is expected to affect the internal pressure.

Katsuyama, Kozo; Ishimi, Akihiro; Maeda, Koji; Nagamine, Tsuyoshi; Asaga, Takeo

2010-06-01

293

Martindale Calculators  

NSDL National Science Digital Library

Martindale Calculators is a Web-based tool collection that contains over 19,000 online calculators created by over "3,450" very "creative" individuals, businesses and â??tax supported entities world wide.â? The collection is organized by the following topics: mathematics; statistics; science A-Z; chemistry; physics, astrophysics and astronomy; engineering A-Z; and electrical engineering, computer engineering, & computer science. Each section includes a wealth of websites to explore, all related to mathematical calculations, mostly course materials and articles. Another section lists online calculators relevant for various industries, such as aviation, cosmetics, insurance, and library science. The list is organized alphabetically and creatively stretches the meaning of â??calculatorâ? to include such things as name translators and databases on animal breeds.

294

Calculating machines  

NSDL National Science Digital Library

This website created by Erez Kaplan "deals mainly with the mechanical calculating machines from a collector's point of view." Included here is an historical review of calculating machines, along with Kaplan's attempt to classify the machines, a collection of old advertisements for the machines, and a brief history of calculating. The latest feature is a Java applet that lets you operate an 1885 Felt adding machine to give you a sense of the way it was used. The photos and descriptions provide insight on other gadgets such as the Pocket Cash Registers used by "the sophisticated man or woman of 1900 who had everything." The Reference section provides some resources for further reading, including numerous other personal calculator collectors sites and museums.

295

Gas core reactors for actinide transmutation and breeder applications  

NASA Technical Reports Server (NTRS)

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

Clement, J. D.; Rust, J. H.

1978-01-01

296

Benchmark data for validating irradiated fuel compositions used in criticality calculations  

SciTech Connect

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.

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

1994-10-01

297

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

NASA Technical Reports Server (NTRS)

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.

Voorhies, Coerte V.

1998-01-01

298

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

NASA Technical Reports Server (NTRS)

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 aide 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 is calculated to be at most 1.5 degrees per year.

Voorhies, C. V.

1999-01-01

299

Model for evolution of crystal defects in UO 2 under irradiation up to high burn-ups  

NASA Astrophysics Data System (ADS)

The model for dislocations evolution under irradiation conditions in UO 2 is developed and implemented in the MFPR code. Being combined with the MFPR set of microscopic equations for the evolution of point defects and their interactions with gas bubbles, a self-consistent consideration of the whole system of point and extended defects in irradiated fuel, including point defects (vacancies, interstitials and gas atoms), as well as extended defects (bubbles, dislocations, vacancy loops and pores), is attained. The MFPR code with the new defect evolution model is successfully validated against steady-irradiation experiments, in which the dislocation density and the bubble concentration and mean size were directly measured as functions of burn-up at ?1000 K. Being applied to higher temperatures, the code allows mechanistic interpretation of the temperature threshold for the fuel restructuring observed in the rim-zone of high burn-up UO 2 fuel.

Veshchunov, M. S.; Shestak, V. E.

2009-01-01

300

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

SciTech Connect

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.

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

301

Curvilinear coordinates for full-core atoms  

NASA Astrophysics Data System (ADS)

Curvilinear coordinates, first introduced by F. Gygi for valence-only electronic systems within the local-density functional theory (F. Gygi, Europhysics Letters 19), 617 (1992)., can be used to describe both core and valence electrons in electronic-structure calculations. A simple and quite general coordinate transformation results in a large, yet affordable plane-wave energy cutoff for full-core systems (e.g., ~= 120 Ryd for carbon or silicon) within the local-density functional theory, and in a reduced correlation time for full-core variational Monte Carlo calculations. Numerical examples will be presented.

Putrino, Anna; Bachelet, Giovanni B.

1998-03-01

302

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

SciTech Connect

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

Broadhead, B.L.

1991-08-01

303

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

Microsoft Academic Search

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

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

2010-01-01

304

Inflation Calculator  

NSDL National Science Digital Library

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.

Friedman, S. Morgan.

305

Mercury Calculator  

NSDL National Science Digital Library

This interactive calculator produced by Teachers' Domain helps you determine the mercury levels in various types of fish, and enables you to make more informed choices about which fish are safe to eat and which should be avoided or eaten infrequently.

2010-09-16

306

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

SciTech Connect

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

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

2008-04-16

307

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

SciTech Connect

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

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

2014-04-01

308

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

SciTech Connect

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.

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

2014-04-01

309

Broken Calculator  

NSDL National Science Digital Library

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.

Mandy Barrow

2008-01-01

310

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

SciTech Connect

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.

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

311

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

SciTech Connect

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.

Jorge Navarro

2013-12-01

312

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

SciTech Connect

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)

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

313

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

NASA Astrophysics Data System (ADS)

Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

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

2013-09-01

314

Chemical Calculations  

NSDL National Science Digital Library

This site contains many chemistry applets created by Jonathan Goodman and his group at Cambridge University. An example of an applet available is the Molecular Weight Calculation; whereby entering in a molecular formula, users are able to discover the HRMS weight, the molecular weight, the element percents, and the Molecular Ion Isotope Pattern. Interactive graphs are also available to assist chemistry students with concepts such as boiling points, pressure, and Consecutive First Step Reversible Reactions. Educators and students will also find many three dimensional depictions of the molecules including fused rings, aromatic rings, and Fullerenes.

Goodman, Jonathan

315

Photoperiod Calculator  

NSDL National Science Digital Library

Jarmo Lammi has developed this simple, easy-to-use tool that provides information useful for teaching and research purposes. Users select a day, month, location (city or latitude and longitude) and time-of-day, and then submit their entry. The Calculator then generates the following information: latitude and longitude for the city/location, declination of the sun, height of sun at noon that day, daylength, and time of sunrise and sunset. This is a useful tool for ecological research and teaching.

Lammi, Jarmo J.

316

An efficient computational technique for light water reactor core dynamics  

Microsoft Academic Search

By combining a modified version of the so-called ''adiabatic'' method for reactor dynamic calculations with a simplified flow redistribution scheme, an efficient method for predicting three-dimensional core behavior has been developed for pressurized water reactor transients. Both the simplified core reactivity and the flow redistribution calculations are shown to yield close approximations of the results obtained by more rigorous approaches.

C. D. Wu; J. Weisman

1988-01-01

317

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

SciTech Connect

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)

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

318

Probability Calculator  

NSDL National Science Digital Library

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.

Stark, Philip B.

319

CALCULATED SECONDARY HEAT GENERATION IN THE D102A REACTOR  

Microsoft Academic Search

The calculated results of secondary heating in the D102A reactor ; assembly are presented. The principal components include the reactor core, front ; and radial reflectors, tube sheets, Inconel pressure vessel, and shield. Heating ; effects considered were neutron moderation, core produced gammas, and extra-core ; neutron capture gammas. Data on distribution of heating among the various ; reactor coraponents

J. G. Keppler; S. R. Lenihan

1958-01-01

320

Academic Rigor: The Core of the Core  

ERIC Educational Resources Information Center

Some educators see the Common Core State Standards as reason for stress, most recognize the positive possibilities associated with them and are willing to make the professional commitment to implementing them so that academic rigor for all students will increase. But business leaders, parents, and the authors of the Common Core are not the only…

Brunner, Judy

2013-01-01

321

Rabacus: Analytic Cosmological Radiative Transfer Calculations  

NASA Astrophysics Data System (ADS)

Rabacus performs analytic radiative transfer calculations in simple geometries relevant to cosmology and astrophysics; it also contains tools to calculate cosmological quantities such as the power spectrum and mass function. With core routines written in Fortran 90 and then wrapped in Python, the execution speed is thousands of times faster than equivalent routines written in pure Python.

Altay, Gabriel

2015-02-01

322

Coring Sample Acquisition Tool  

NASA Technical Reports Server (NTRS)

A sample acquisition tool (SAT) has been developed that can be used autonomously to sample drill and capture rock cores. The tool is designed to accommodate core transfer using a sample tube to the IMSAH (integrated Mars sample acquisition and handling) SHEC (sample handling, encapsulation, and containerization) without ever touching the pristine core sample in the transfer process.

Haddad, Nicolas E.; Murray, Saben D.; Walkemeyer, Phillip E.; Badescu, Mircea; Sherrit, Stewart; Bao, Xiaoqi; Kriechbaum, Kristopher L.; Richardson, Megan; Klein, Kerry J.

2012-01-01

323

The effects of irradiation condition and microstructural change on lattice parameter, crystal lattice strain and crystallite size in high burnup UO 2 pellet  

NASA Astrophysics Data System (ADS)

Pellet samples (average burnup: 37-62 GWd/t) were prepared from three kinds of fuel rods which were irradiated in the Halden Heavy Water Reactor in Norway, and microstructural changes in the pellet samples were investigated by means of optical microscopy, SEM/EPMA and micro-X-ray diffractometry. The measured lattice parameters tended to be smaller than the values reported previously, and it is likely that the measured lattice parameters were affected by the temperature conditions during the irradiation tests and the microstructural changes which occurred in the samples. Considering the burnup dependence of uniform and non-uniform strains, the following things are suggested: the interstitial atoms which cause uniform strain begin firstly to form dislocations as a recovery process of irradiation defect and the dislocation density increases. With increasing burnup, the accumulation of dislocations in the crystallite saturates and the migration of dislocations becomes dominant as a recovery process of irradiation defects in the crystallite.

Amaya, Masaki; Nakamura, Jinichi; Fuketa, Toyoshi

2009-08-01

324

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

SciTech Connect

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.

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

2008-10-24

325

Banded transformer cores  

NASA Technical Reports Server (NTRS)

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.

Mclyman, C. W. T. (inventor)

1974-01-01

326

Reactor whole core transport calculations without fuel assembly homogenization  

SciTech Connect

The variational nodal method is generalized by dividing each spatial node into a number of triangular finite elements designated as subelements. The finite subelement trail functions allow for explicit geometry representations within each node, thus eliminating the need for nodal homogenization. The method is implemented within the Argonne National Laboratory code VARIANT and applied to two-dimensional multigroup problems. Eigenvalue and pin-power results are presented for a four-assembly OECD/NEA benchmark problem containing enriched U{sub 2} and MOX fuel pins. Our seven-group model combines spherical or simplified spherical harmonic approximations in angle with isoparametric linear or quadratic subelement basis functions, thus eliminating the need for fuel-coolant homogenization. Comparisons with reference seven-group Monte Carlo solutions indicate that in the absence of pin-cell homogenization, high-order angular approximations are required to obtain accurate eigenvalues, while the results are substantially less sensitive to the refinement of the finite subelement grids.

Nicholas Tsoulfanidis; Elmer Lewis; M.A. Smith; G. Palmiotti; T.A. Taiwo

2002-10-18

327

Inner core tilt and polar motion  

NASA Astrophysics Data System (ADS)

A tilted inner core permits exchange of angular momentum between the core and the mantle through gravitational and pressure torques and, as a result, changes in the direction of Earth's axis of rotation with respect to the mantle. We have developed a model to calculate the amplitude of the polar motion that results from an equatorial torque at the inner core boundary which tilts the inner core out of alignment with the mantle. We specifically address the issue of the role of the inner core tilt in the decade polar motion known as the Markowitz wobble. We show that a decade polar motion of the same amplitude as the observed Markowitz wobble requires a torque of 1020 N m which tilts the inner core by 0.07 degrees. This result critically depends on the viscosity of the inner core; for a viscosity less than 5 × 1017 Pa s, larger torques are required. We investigate the possibility that a torque of 1020 N m with decadal periodicity can be produced by electromagnetic coupling between the inner core and torsional oscillations of the flow in the outer core. We demonstrate that a radial magnetic field at the inner core boundary of 3 to 4 mT is required to obtain a torque of such amplitude. The resulting polar motion is eccentric and polarized, in agreement with the observations. Our model suggests that equatorial torques at the inner core boundary might also excite the Chandler wobble, provided there exists a physical mechanism that can generate a large torque at a 14 month period.

Dumberry, Mathieu; Bloxham, Jeremy

2002-11-01

328

HYDRATE CORE DRILLING TESTS  

SciTech Connect

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.

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

2002-11-01

329

Core-Cutoff Tool  

NASA Technical Reports Server (NTRS)

A tool makes a cut perpendicular to the cylindrical axis of a core hole at a predetermined depth to free the core at that depth. The tool does not damage the surrounding material from which the core was cut, and it operates within the core-hole kerf. Coring usually begins with use of a hole saw or a hollow cylindrical abrasive cutting tool to make an annular hole that leaves the core (sometimes called the plug ) in place. In this approach to coring as practiced heretofore, the core is removed forcibly in a manner chosen to shear the core, preferably at or near the greatest depth of the core hole. Unfortunately, such forcible removal often damages both the core and the surrounding material (see Figure 1). In an alternative prior approach, especially applicable to toxic or fragile material, a core is formed and freed by means of milling operations that generate much material waste. In contrast, the present tool eliminates the damage associated with the hole-saw approach and reduces the extent of milling operations (and, hence, reduces the waste) associated with the milling approach. The present tool (see Figure 2) includes an inner sleeve and an outer sleeve and resembles the hollow cylindrical tool used to cut the core hole. The sleeves are thin enough that this tool fits within the kerf of the core hole. The inner sleeve is attached to a shaft that, in turn, can be attached to a drill motor or handle for turning the tool. This tool also includes a cutting wire attached to the distal ends of both sleeves. The cutting wire is long enough that with sufficient relative rotation of the inner and outer sleeves, the wire can cut all the way to the center of the core. The tool is inserted in the kerf until its distal end is seated at the full depth. The inner sleeve is then turned. During turning, frictional drag on the outer core pulls the cutting wire into contact with the core. The cutting force of the wire against the core increases with the tension in the wire and, hence, with the frictional drag acting on the outer sleeve. As the wire cuts toward the center of the core, the inner sleeve rotates farther with respect to the outer sleeve. Once the wire has cut to the center of the core, the tool and the core can be removed from the hole. The proper choice of cutting wire depends on the properties of the core material. For a sufficiently soft core material, a nonmetallic monofilament can be used. For a rubber-like core material, a metal wire can be used. For a harder core material, it is necessary to use an abrasive wire, and the efficiency of the tool can be increased greatly by vacuuming away the particles generated during cutting. For a core material that can readily be melted or otherwise cut by use of heat, it could be preferable to use an electrically heated cutting wire. In such a case, electric current can be supplied to the cutting wire, from an electrically isolated source, via rotating contact rings mounted on the sleeves.

Gheen, Darrell

2007-01-01

330

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

SciTech Connect

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.

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

331

Core Coupling in Nb-99  

E-print Network

to be good. NUCLEAR REACTIONS, NUCLEAR STRUCTURE '"Mop, 'He), E =40.7 Mev, measured o(8) 99Nb levels deduced S. Calculated levels, 4, n', 8 NNb, particle- core-coupling model. The proton configurations of nuclei in the Zr-Mo region have been the subject... to be 0.47, 0.76, and 1.46 MeV for the 2P, ~?2P,~?and 1f,~, orbitals, respective- ly, and were not varied. The Hamiltonian matrices for I= ~, &, & .. and ~2 were diagonalized to obtain the energy eigenvalues and eigenfunctions for values of $, and X...

Bindal, P. K.; Youngblood, David H.

1974-01-01

332

Core sample extractor  

NASA Technical Reports Server (NTRS)

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.

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

1989-01-01

333

Dublin Core Metadata Initiative  

NSDL National Science Digital Library

Dublin Core metadata has been implemented in several ways, including as HTML metatags and as database elements, as it is used in the Scout Archives (discussed in the June 20, 1997 issue of the Scout Report). The DC elements are title, author, subject, description, publisher, other contributor, date, resource type, format, resource identifier, source, language, relation, coverage, and rights management. Information about the Dublin Core Workshop Series, DC semantics and syntax, working papers, and projects that have implemented Dublin Core metadata can be found at the Dublin Core Metadata homepage.

334

KSI's Cross Insulated Core Transformer Technology  

SciTech Connect

Cross Insulated Core Transformer (CCT) technology improves on Insulated Core Transformer (ICT) implementations. ICT systems are widely used in very high voltage, high power, power supply systems. In an ICT transformer ferrite core sections are insulated from their neighboring ferrite cores. Flux leakage is present at each of these insulated gaps. The flux loss is raised to the power of stages in the ICT design causing output voltage efficiency to taper off with increasing stages. KSI's CCT technology utilizes a patented technique to compensate the flux loss at each stage of an ICT system. Design equations to calculate the flux compensation capacitor value are presented. CCT provides corona free operation of the HV stack. KSI's CCT based High Voltage power supply systems offer high efficiency operation, high frequency switching, low stored energy and smaller size over comparable ICT systems.

Uhmeyer, Uwe [Kaiser Systems, Inc, 126 Sohier Road, Beverly, MA 01915 (United States)

2009-08-04

335

KSI's Cross Insulated Core Transformer Technology  

NASA Astrophysics Data System (ADS)

Cross Insulated Core Transformer (CCT) technology improves on Insulated Core Transformer (ICT) implementations. ICT systems are widely used in very high voltage, high power, power supply systems. In an ICT transformer ferrite core sections are insulated from their neighboring ferrite cores. Flux leakage is present at each of these insulated gaps. The flux loss is raised to the power of stages in the ICT design causing output voltage efficiency to taper off with increasing stages. KSI's CCT technology utilizes a patented technique to compensate the flux loss at each stage of an ICT system. Design equations to calculate the flux compensation capacitor value are presented. CCT provides corona free operation of the HV stack. KSI's CCT based High Voltage power supply systems offer high efficiency operation, high frequency switching, low stored energy and smaller size over comparable ICT systems.

Uhmeyer, Uwe

2009-08-01

336

VLBI Observations of the Free Core Nutations  

NASA Astrophysics Data System (ADS)

At core scale lengths with periods from a few hours to days, the Coriolis acceleration dominates the Lorentz force density and core modes can be considered as purely mechanical. One of the most interesting core modes is the spin-over mode, which reflects the ability of the outer core to rotate about an axis different from that of either the inner core or the shell. It has a nearly diurnal period. In the Earth frame of reference, this mode produces the nearly diurnal retrograde wobble. In the space frame of reference it is accompanied by the free core nutations. When the flattening of the boundaries of the fluid outer core and the figure-figure gravitational coupling are taken into account, as well as the deformability of the boundaries, both a retrograde free core nutation and a prograde free core nutation are found. The retrograde free core nutation was first predicted by Poincare (1910) for a completly fluid, incompressible core bounded by a rigid shell. In a variational calculation of wobble-nutation modes in realistic Earth models, Jiang (1993) found the classical retrograde free core nutation (RFCN) but discovered a prograde free core nutation (PFCN) as well. VLBI residuals in longitude and obliquity compared to the 1980 IAU nutation series, and their standard errors, were downloaded from the Goddard Space Flight Center website, for the period August 3, 1979 to March 6, 2003, giving 3343 points over a span of 8617 days. In an overlapping segment analysis, the discrete Fourier transform (DFT) for each segment was found for the corresponding series of unequally spaced nutation residuals by singular value decomposition (SVD), with the number of singular values eliminated determined by the satisfaction of Parseval's theorem. Both the RFCN and the PFCN resonances were found in the resulting power spectrum. The nutation resonances were found to be in free decay, the half-life of the PFCN at 2620 days and that of the RFCN at 2229 days, with Ekman boundary layer theory leading to viscosities at the top of the core of 3124 Pa s and 3611 Pa s, respectively.

Smylie, D. E.

2012-12-01

337

On the rate determining step in fission gas release from high burn-up water reactor fuel during power transients  

NASA Astrophysics Data System (ADS)

The radial distribution of grain boundary gas in a PWR and a BWR fuel is reported. The measurements were made using a new approach involving X-ray fluorescence analysis and electron probe microanalysis. In both fuels the concentration of grain boundary gas was much higher than hitherto suspected. The gas was mainly contained in the bubble/pore structure. The factors that determined the fraction of gas released from the grains and the level of gas retention on the grain boundaries are identified and discussed. The variables involved are the local fuel stoichiometry, the amount of open porosity, the magnitude of the local compressive hydrostatic stress and the interaction of metallic precipitates with gas bubbles on the grain faces. It is concluded that under transient conditions the interlinkage of gas bubbles on the grain faces and the subsequent formation of grain edge tunnels is the rate determining step for gas release; at least when high burn-up fuel is involved.

Walker, C. T.; Mogensen, M.

1987-07-01

338

MOX capsule post-irradiation examination. Volume 2: Test plan for 30-GWd/MT burnup fuel  

SciTech Connect

This test plan is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. The planned post-irradiation examination (PIE) work to be performed on the mixed uranium and plutonium oxide fuel capsules that have received burnups of approximately 30 GWd/MT is described. The major emphasis of this PIE task will be material interactions, particularly indications of gallium transport and interactions. This PIE will include gamma scanning, ceramography, metallography, pellet radial gallium analysis, and clad gallium analysis. A preliminary PIE report will be generated before all the work is completed so that the progress of the fuel irradiation may be known in a timely manner.

Morris, R.N.

1997-12-01

339

A Technique to Determine Billet Core Charge Weight for P/M Fuel Tubes  

SciTech Connect

The core length in an extruded tube depends on the weight of powder in the billet core. In the past, the amount of aluminum powder needed to give a specified core length was determined empirically. This report gives a technique for calculating the weight of aluminum powder for the P/M core. An equation has been derived which can be used to determine the amount of aluminum needed for P/M billet core charge weights. Good agreement was obtained when compared to Mark 22 tube extrusion data. From the calculated charge weight, the elastomeric bag can be designed and made to compact the U3O8-Al core.

Peacock, H.B.

2001-07-02

340

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

341

NFE Core Bibliographies.  

ERIC Educational Resources Information Center

This collection of core bibliographies, which expands on an initial bibliography published in 1979 of the core resources housed in the Non-Formal Education Information Center at Michigan State University, comprises a basic stock of materials on nonformal education and women in development that have been contributed by development planners,…

Michigan State Univ., East Lansing. Inst. for International Studies in Education.

342

Making an Ice Core.  

ERIC Educational Resources Information Center

Explains an activity in which students construct a simulated ice core. Materials required include only a freezer, food coloring, a bottle, and water. This hands-on exercise demonstrates how a glacier is formed, how ice cores are studied, and the nature of precision and accuracy in measurement. Suitable for grades three through eight. (Author/PVD)

Kopaska-Merkel, David C.

1995-01-01

343

Ice Core Investigations  

ERIC Educational Resources Information Center

What can glaciers tell us about volcanoes and atmospheric conditions? How does this information relate to our understanding of climate change? Ice Core Investigations is an original and innovative activity that explores these types of questions. It brings together popular science issues such as research, climate change, ice core drilling, and air…

Krim, Jessica; Brody, Michael

2008-01-01

344

Cores of convex games  

Microsoft Academic Search

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.

Lloyd S. Shapley

1971-01-01

345

Core Concepts of Kinesiology.  

ERIC Educational Resources Information Center

Core concepts of kinesiology are the basis of communication about movement that facilitate progression of skill levels. The article defines and exemplifies each of 10 core concepts: range of motion, speed of motion, number of segments, nature of segments, balance, coordination, compactness, extension at release/contact, path of projection, and…

Hudson, Jackie L.

1995-01-01

346

Reading Antarctica's Rock Cores  

NSDL National Science Digital Library

In this activity, students learn about the tools and methods paleoclimatologists use to reconstruct past climates. In constructing sediment cores themselves, students will achieve a very good understanding of the sedimentological interpretation of past climates that scientists can draw from cores.

LuAnn Dahlman

347

Reactivity Coefficients in BN600 Core with Minor Actinides  

Microsoft Academic Search

In 1999, the IAEA has initiated a Coordinated Research Project on “Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects.” Three benchmark models representing different modifications of the BN-600 fast reactor have been sequentially established and analyzed, including a hybrid core with highly enriched uranium oxide and MOX fuel, a full MOX core with weapons-grade

Andrei RINEISKI; Makoto ISHIKAWA; Jinwook JANG; Prabhakaran MOHANAKRISHNAN; Tim NEWTON; Gérald RIMPAULT; Alexander STANCULESCU; Victor STOGOV

2011-01-01

348

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...

349

Lunar Core and Tides  

NASA Technical Reports Server (NTRS)

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.

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

2004-01-01

350

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

351

Core drill and method of removing a core therefrom  

Microsoft Academic Search

This patent describes a method of expediting the removal of a core from the interior of a tubular core drill which comprises: fixedly securing an externally threaded bushing to the rear end of the core drill; providing a sleeve for detachably coupling the bushing-equipped core drill to the externally threaded drive shank of a power unit for the core drill.

Bossler

1987-01-01

352

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

NASA Astrophysics Data System (ADS)

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.

Yoo, Myoung-Woo; Kim, Sang-Koog

2015-01-01

353

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

SciTech Connect

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.

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

2008-06-01

354

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

NASA Technical Reports Server (NTRS)

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.

Boss, Alan P.

1993-01-01

355

Iron diffusion from first principles calculations  

NASA Astrophysics Data System (ADS)

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).

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

2013-12-01

356

Chemical Models of Star-Forming Cores  

NASA Astrophysics Data System (ADS)

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.

Aikawa, Y.

2013-10-01

357

HENRY'S LAW CALCULATOR  

EPA Science Inventory

On-Site was developed to provide modelers and model reviewers with prepackaged tools ("calculators") for performing site assessment calculations. The philosophy behind OnSite is that the convenience of the prepackaged calculators helps provide consistency for simple calculations,...

358

Reevaluation of Assay Data of Spent Nuclear Fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems  

Microsoft Academic Search

The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent

Minoru Murazaki; Kiyoshi Ohkubo; Yoshinori Nakahara; Gunzo Uchiyama

2011-01-01

359

Statistics of X-ray observables for the cooling-core and non-cooling core galaxy clusters  

E-print Network

We present a statistical study of the occurrence and effects of the cooling cores in the clusters of galaxies in a flux-limited sample, HIFLUGCS, based on ROSAT and ASCA observations. About 49% of the clusters in this sample have a significant, classically-calculated cooling-flow, mass-deposition rate. The upper envelope of the derived mass-deposition rate is roughly proportional to the cluster mass, and the fraction of cooling core clusters is found to decrease with it. The cooling core clusters are found to have smaller core radii than non-cooling core clusters, while some non-cooling core clusters have high $\\beta$ values (> 0.8). In the relation of the X-ray luminosity vs. the temperature and the mass, the cooling core clusters show a significantly higher normalization. A systematic correlation analysis, also involving relations of the gas mass and the total infrared luminosity, indicates that this bias is shown to be mostly due to an enhanced X-ray luminosity for cooling core clusters, while the other parameters, like temperature, mass, and gas mass may be less affected by the occurrence of a cooling core. These results may be explained by at least some of the non-cooling core clusters being in dynamically young states compared with cooling core clusters, and they may turn into cooling core clusters in a later evolutionary stage.

Y. Chen; T. H. Reiprich; H. Böhringer; Y. Ikebe; Y. -Y. Zhang

2007-02-19

360

Wireline sidewall coring  

SciTech Connect

In April 1989, Schlumberger Well Services, under contract to Fenix and Scisson of Nevada, Inc., ran a wireline sidewall coring machine in exploratory hole Ue4t at the Nevada Test Site for the Los Alamos National Laboratory. The sampling project goals were to recover material for geologic characterization and to determine the effectiveness of the tool for sampling various volcanic lithologies. If a wireline tool is found to be effective, fewer expensive continuously-cored holes will be needed. The Schlumberger Sidewall Coredriller has a maximum diameter of 5.25 inches and, with the gamma-ray unit included for stratigraphic correlation, is approximately 40 feet long. It weighs 850 pounds. All the downhole mechanical systems are hydraulic including the anchor shoe, the coring motor, the pressure on the bit and the core extraction system. Sonde functions are monitored and controlled at the surface. The tool is designed to run in fluid with the waterways in the diamond but creating circulation to keep the bit face clean. Up to 20 cores, measuring 0.91 inches in diameter by 2 inches long, can be recovered with each each. These cores are separated in the split-sleeve catcher tube by discs automatically inserted following each coring. 1 ref., 4 figs., 1 tab.

Hawkins, W.L.; Mathews, M.A. (Los Alamos National Lab., NM (USA)); Thompson, P.H. (Fenix and Scisson, Inc., Mercury, NV (USA)); Jenkins, K. (Schlumberger Well Services, Casper, WY (USA))

1989-01-01

361

Analysis of fresh fuel critical experiments appropriate for burnup credit validation  

SciTech Connect

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.

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

1995-10-01

362

The Expanded FindCore Method for Identification of a Core Atom Set for Assessment of Protein Structure Prediction  

PubMed Central

Maximizing the scientific impact of NMR-based structure determination requires robust and statistically sound methods for assessing the precision of NMR-derived structures. In particular, a method to define a core atom set for calculating superimpositions and validating structure predictions is critical to the use of NMR-derived structures as targets in the CASP competition. FindCore (D.A. Snyder and G.T. Montelione PROTEINS 2005;59:673–686) is a superimposition independent method for identifying a core atom set, and partitioning that set into domains. However, as FindCore optimizes superimposition by sensitively excluding not-well-defined atoms, the FindCore core may not comprise all atoms suitable for use in certain applications of NMR structures, including the CASP assessment process. Adapting the FindCore approach to assess predicted models against experimental NMR structures in CASP10 required modification of the FindCore method. This paper describes conventions and a standard protocol to calculate an “Expanded FindCore” atom set suitable for validation and application in biological and biophysical contexts. A key application of the Expanded FindCore method is to identify a core set of atoms in the experimental NMR structure for which it makes sense to validate predicted protein structure models. We demonstrate the application of this Expanded FindCore method in characterizing well-defined regions of 18 NMR-derived CASP10 target structures. The Expanded FindCore protocol defines “expanded core atom sets” that match an expert’s intuition of which parts of the structure are sufficiently well-defined to use in assessing CASP model predictions. We also illustrate the impact of this analysis on the CASP GDT assessment scores. PMID:24327305

Snyder, David A.; Grullon, Jennifer; Huang, Yuanpeng J.; Tejero, Roberto; Montelione, Gaetano T.

2014-01-01

363

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.

364

Core assembly storage structure  

DOEpatents

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.

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

1988-01-01

365

Core helium flash  

SciTech Connect

The role of convection in the core helium flash is simulated by two-dimensional eddies interacting with the thermonuclear runaway. These eddies are followed by the explicit solution of the 2D conservation laws with a 2D finite difference hydrodynamics code. Thus, no phenomenological theory of convection such as the local mixing length theory is required. The core helium flash is violent, producing a deflagration wave. This differs from the detonation wave (and subsequent disruption of the entire star) produced in previous spherically symmetric violent core helium flashes as the second dimension provides a degree of relief which allows the expansion wave to decouple itself from the burning front. Our results predict that a considerable amount of helium in the core will be burned before the horizontal branch is reached and that some envelope mass loss is likely.

Cole, P.W.; Deupree, R.G.

1980-01-01

366

Core Manager: Ellen Sisk  

NSDL National Science Digital Library

This is a PDF interview, PowerPoint slide set, and webpage biography of a core manager, detailing the importance of a lab manager to oversee the complex workings of DNA sequencing machines for an entire company.

2012-05-02

367

Micro coring apparatus  

NASA Technical Reports Server (NTRS)

A micro-coring apparatus for lunar exploration applications, that is compatible with the other components of the Walking Mobile Platform, was designed. The primary purpose of core sampling is to gain an understanding of the geological composition and properties of the prescribed environment. This procedure has been used extensively for Earth studies and in limited applications during lunar explorations. The corer is described and analyzed for effectiveness.

Collins, David; Brooks, Marshall; Chen, Paul; Dwelle, Paul; Fischer, Ben

1989-01-01

368

Central core disease  

Microsoft Academic Search

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;

Heinz Jungbluth

2007-01-01

369

Fe-based nanocrystalline FINEMET cores for induction accelerators  

NASA Astrophysics Data System (ADS)

Pulse power modulators used for high repetition rate linear induction accelerators are generally composed of main switches using thyratrons, step-up pulse transformers, multistage magnetic pulse compression circuits and pulse forming lines. Fe-based amorphous cores are normally used for saturable reactors in linear induction accelerators, because the volume of the saturable reactor using Fe-based cores is small due to its high effective induction swing. However, there is a great demand to reduce the core loss in order to improve the efficiency of pulse power modulators. Recently, a new Fe-based nanocrystalline soft magnetic materials, FINEMET, that has both high saturation flux density and low core loss was developed. The dynamic magnetic characteristics of an insulated FINEMET core were measured using a high voltage pulse excitation circuit, and the total core volume and the total core loss for saturable reactors in a 2-stage magnetic pulse compression circuit were calculated. As the result, by using the FINEMET cores for saturable reactors in a magnetic pulse compression circuit, the total core loss becomes 68%, and the total core volume becomes 133%, of the conventional Fe-based amorphous cores.

Nakajima, S.; Arakawa, S.; Yamashita, Y.; Shiho, M.

1993-07-01

370

MCNP LWR Core Generator  

SciTech Connect

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.

Fischer, Noah A. [Los Alamos National Laboratory

2012-08-14

371

Track "core" effects in heavy ion radiolysis  

NASA Astrophysics Data System (ADS)

By assuming that HO .2 radical production in water and H 2 production in benzene are 2 hit processes, and applying the concepts of track physics, we are able to obtain a parameteric fit to the yields of these reactions by heavy ion radiolysis from knowledge of the radial dose distribution about a heavy ion's path. We make no use of the concept of a track core, for no clearly definable track core appears in our calculations of the radial dose distribution. Instead we calculate an action cross section ? from the assumed 2 hit response to ?-rays. The cross section is calculated from two fitted parameters, E0, the ?-ray dose at which there is an average of 1 hit per target, and the target radius a0. From the cross section, the target radius and the stopping power we calculate the G value. While our model is not mechanistic, the assumed 2 hit process is consistent with hypotheses which have been offered as chemical models for these processes. Since a 2 hit process is more likely to take place in a high dose region, close to an ion's path, it may easily be attributed to a hypothetical track core in energy deposition, when indeed the response is a property of the detector.

Katz, Robert; Huang, Guo-Rong

372

Plasma core at the center of a sonoluminescing bubble  

NASA Astrophysics Data System (ADS)

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.

Bemani, F.; Sadighi-Bonabi, R.

2013-01-01

373

Plasma core at the center of a sonoluminescing bubble.  

PubMed

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

Bemani, F; Sadighi-Bonabi, R

2013-01-01

374

Vintage Calculators Web Museum  

NSDL National Science Digital Library

This "web museum" devoted to vintage calculators shows "the evolution from mechanical calculator to hand held electronic calculator." Some items featured include: Mechanical and early electronic desk calculators, "strange hand-held calculators," and articles, photographs, and databases from the archives of the International Association of Calculator Collectors. A history of the technology and information on British and sterling currency calculators are also posted here. The website also offers a Calculator time-line (chronicling calculator developments), background on the technology used by mechanical and early electronic calculators, and information on The Calculator Business. An index allows visitors to search the calculators featured on this site. The Puzzle Corner section asks visitors to contact them with any information that may answer unresolved questions regarding vintage calculators.

Tout, Nigel

375

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.

376

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.

377

Core-Noise  

NASA Technical Reports Server (NTRS)

This presentation is a technical progress report and near-term outlook for NASA-internal and NASA-sponsored external work 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; the emerging importance of core noise and its relevance to the SFW Reduced-Noise-Aircraft Technical Challenge; the current research activities in the core-noise area, with some additional details given about the development of a high-fidelity combustion-noise prediction capability; the need for a core-noise diagnostic capability to generate benchmark data for validation of both high-fidelity work and improved models, as well as testing of future noise-reduction technologies; relevant existing core-noise tests using real engines and auxiliary power units; and examples of possible scenarios for a future diagnostic facility. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Reduced-Noise-Aircraft Technical Challenge aims to enable concepts and technologies to dramatically reduce the perceived aircraft noise outside of airport boundaries. This reduction of aircraft noise is critical for enabling the anticipated large increase in future air traffic. Noise generated in the jet engine core, by sources such as the compressor, combustor, and turbine, can be a significant contribution to the overall noise signature at low-power conditions, typical of approach flight. At high engine power during takeoff, jet and fan noise have traditionally dominated over core noise. However, current design trends and expected technological advances in engine-cycle design as well as noise-reduction methods are likely to reduce non-core noise even at engine-power points higher than approach. In addition, future low-emission combustor designs could increase the combustion-noise component. The trend towards high-power-density cores also means that the noise generated in the low-pressure turbine will likely increase. Consequently, the combined result from these emerging changes will be to elevate the overall importance of turbomachinery core noise, which will need to be addressed in order to meet future noise goals.

Hultgren, Lennart S.

2010-01-01

378

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)

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.

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

379

Core Noise - Increasing Importance  

NASA Technical Reports Server (NTRS)

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 Reduced-Perceived-Noise Technical Challenge; and the current research activities in the core-noise area, with additional details given about the development of a high-fidelity combustor-noise prediction capability as well as activities supporting the development of improved reduced-order, physics-based models for combustor-noise prediction. The need for benchmark data for validation of high-fidelity and modeling work and the value of a potential future diagnostic facility for testing of core-noise-reduction concepts are indicated. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Reduced-Perceived-Noise Technical Challenge aims to develop concepts and technologies to dramatically reduce the perceived aircraft noise outside of airport boundaries. This reduction of aircraft noise is critical to enabling the anticipated large increase in future air traffic. Noise generated in the jet engine core, by sources such as the compressor, combustor, and turbine, can be a significant contribution to the overall noise signature at low-power conditions, typical of approach flight. At high engine power during takeoff, jet and fan noise have traditionally dominated over core noise. However, current design trends and expected technological advances in engine-cycle design as well as noise-reduction methods are likely to reduce non-core noise even at engine-power points higher than approach. In addition, future low-emission combustor designs could increase the combustion-noise component. The trend towards high-power-density cores also means that the noise generated in the low-pressure turbine will likely increase. Consequently, the combined result from these emerging changes will be to elevate the overall importance of turbomachinery core noise, which will need to be addressed in order to meet future noise goals.

Hultgren, Lennart S.

2011-01-01

380

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

SciTech Connect

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)

Loida, Andreas; Metz, Volker; Kienzler, Bernhard [Institut fuer Nukleare Entsorgung, Forschungszentrum Karlsruhe, P.O.Box 3640, Karlsruhe, D- 76021 (Germany)

2007-07-01

381

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

NASA Astrophysics Data System (ADS)

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.

Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji

2011-09-01

382

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

SciTech Connect

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.

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

2004-10-01

383

Advanced core design and fuel management for pebble-bed reactors  

NASA Astrophysics Data System (ADS)

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.

Gougar, Hans David

384

Core Noise Reduction  

NASA Technical Reports Server (NTRS)

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.

Hultgren, Lennart S.

2011-01-01

385

Does Mercury have a molten core  

NASA Technical Reports Server (NTRS)

The question of whether or not Mercury could contain a molten metallic core is investigated by studying the possible thermal evolution of a metallic core in that planet. The calculations involve the solution of the equation of heat conduction for a spherically symmetric body with internal heat sources, modifications to take account of the latent heat of fusion as well as the redistribution of radioactive heat sources as a consequence of melting, the terrestrial Fe/U ratio, and a Th/U ratio of 3.7. The temperature profile predicted by the calculations for a period of 4.6 billion years indicates that the inner 1400 km of the core would now be solid while the outer 500 km would be molten. It is emphasized that this result is a direct consequence of a discontinuity in melting temperatures at the core-mantle boundary and that although a dynamo is possible, it would have to be driven mechanically rather than by thermal convection.

Fricker, P. E.; Reynolds, R. T.; Summers, A. L.; Cassen, P. M.

1976-01-01

386

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis  

SciTech Connect

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

Basehore, K.L.; Todreas, N.E.

1980-08-01

387

Core Training: Stabilizing the Confusion  

Microsoft Academic Search

summary Confusion exists regarding what the core musculature is, how it is evalu- ated, how it is trained, and how it is applied to functional performance. The core musculature is divided into 2 systems, local (stabilization) and global (movement), with distinction between core-strength, core-stabili- ty, and functional exercises.

Mark D. Faries; Mike Greenwood

2007-01-01

388

Distillation Calculations with a Programmable Calculator.  

ERIC Educational Resources Information Center

Describes a three-step approach for teaching multicomponent distillation to undergraduates, emphasizing patterns of distribution as an aid to understanding the separation processes. Indicates that the second step can be carried out by programmable calculators. (A more complete set of programs for additional calculations is available from the…

Walker, Charles A.; Halpern, Bret L.

1983-01-01

389

Core Principles Methodology  

NSDL National Science Digital Library

This newly published document from the Basel Committee on Banking Supervision at the Bank of International Settlements considers the methodology used in determining The Core Principles for Effective Banking Supervision, "a global standard for prudential regulation and supervision," which has been endorsed by many countries worldwide. There are three sections to the report. The first chapter looks at the background for the core principles and "the preconditions for effective banking supervision." The second chapter "raises a few basic considerations regarding the conduct of an assessment and the compilation and presentation of the results," and the last chapter discusses each core principle individually. The 56-page document is available in .pdf format. A thumbnail map of each page, shown on the left, is the best way to navigate the report.

390

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

NASA Astrophysics Data System (ADS)

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

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

2014-12-01

391

Thermal mass limit of neutron cores  

NASA Astrophysics Data System (ADS)

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 .

Roupas, Zacharias

2015-01-01

392

CFD Analysis of Core Bypass Phenomena  

SciTech Connect

The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary

Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

2009-11-01

393

CFD Analysis of Core Bypass Phenomena  

SciTech Connect

The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the sector grid can be set as a symmetry boundary

Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

2010-03-01

394

Calculations of radiation levels during reactor operations for safeguard inspections  

NASA Astrophysics Data System (ADS)

It is necessary to calculate the total level of radioactivity produced by the reactor operation when the reactor cover is opened during periodical safeguard inspections for safety requirements. Also the calculations are performed for different reactor core and spent fuel storage loadings during refuelling after reactor shutdown. When an internal core spent fuel storage is used in the shield tank to accommodate a large number of spent fuel baskets, physical calculations are performed to determine the number of these spent fuel elements which can be accommodated and still maintain the gamma activity outside under the permissible limit. The corresponding reactor power level is determined. The radioactivity calculations are performed for this internal storage at different axial levels to avoid the criticality of the reactor core. Transport theory is used in calculations based on collision probability for multi group cell calculations. Diffusion theory is used in three dimensions in the core calculations. The nuclear fuel history is traced and radioactive decay is calculated, since reactor fission products are very sensitive to power level. The radioactivity is calculated with a developed formula based on fuel basket loading integrity.

Sobhy, M.

1996-05-01

395

Ice Core Exercise  

NSDL National Science Digital Library

Students access the ice core data archived at Lamont-Doherty Geological Observatory. They select a core (Greenland, Antarctica, Quelcaya), pose a working hypothesis regarding the data, import the data in an Excel-readable format, and examine the data to determine correlations between variables and cause/effect as recorded in leads and lags. They generate a written and graphical analysis of the data and, in the next lab period, discuss the similarities and differences among their group outputs in terms of demonstrated correlations, assumptions required, effects of latitude, and any other item that arises.

William Locke

396

Shear viscosity in neutron star cores  

E-print Network

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.

P. S. Shternin; D. G. Yakovlev

2008-08-21

397

Autistic Savant Calendar Calculators.  

ERIC Educational Resources Information Center

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…

Patti, Paul J.

398

Benchmarking of the FIBWR2 transient BWR (boiling water reactor) core hydraulics code  

Microsoft Academic Search

The FIBWR code has been widely used for boiling water reactor (BWR) steady-state core flow, pressure, and void distribution calculations. The code models the complex flow paths in a BWR core, including water tubes and leakage flow to the bypass. FIBWR results are used to calculate flow-weighted equivalent loss coefficients for geometrically more simplified codes such as SIMULATE and RETRAN.

B. J. Gitnick; D. A. Prelewicz

1990-01-01

399

Neutrino-nucleon reaction rates in the supernova core in the relativistic random phase approximation  

Microsoft Academic Search

We calculate neutrino reaction rates with nucleons via the neutral and charged currents in the supernova core in the relativistic random phase approximation (RPA) and study their effects on the opacity of the supernova core. The formulation is based on the Lagrangian employed in the calculation of the nuclear equation of state (EOS) in the relativistic mean field theory (RMF).

Shoichi Yamada; Hiroshi Toki

2000-01-01

400

BWR AXIAL PROFILE  

SciTech Connect

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.

J. Huffer

2004-09-28

401

Voice over IP Calculator  

NSDL National Science Digital Library

The Voice over IP Calculator Web site actually consists of four free online tools that can be used to estimate bandwidth requirements and voice paths for a planned VoIP system. The four tools are: Lines and IP Bandwidth Calculator, Erlangs and Bandwidth Calculator, Minutes and Lines Calculator, and Erlangs and Lines Calculator. Each utility is very easy to use, but is mainly intended for experienced IT workers.

402

Reconstruction of frozen-core all-electron orbitals from pseudo-orbitals  

Microsoft Academic Search

We investigate the numerical feasibility of reconstructing frozen-core all-electron molecular orbitals from corresponding pseudo-orbitals. We perform density-functional calculations on simple atomic and molecular model systems using ultrasoft pseudopotentials to represent the atomic cores. We apply a transformation due to Blo¨chl [Phys. Rev. B 50, 17953 (1994)] to each calculated pseudo-orbital to obtain a corresponding frozen-core all-electron molecular orbital. Our model

Bala´zs Hete´nyi; Filippo De Angelis; Paolo Giannozzi; Roberto Car

2001-01-01

403

Reconstruction of frozen-core all-electron orbitals from pseudo-orbitals  

Microsoft Academic Search

We investigate the numerical feasibility of reconstructing frozen-core all-electron molecular orbitals from corresponding pseudo-orbitals. We perform density-functional calculations on simple atomic and molecular model systems using ultrasoft pseudopotentials to represent the atomic cores. We apply a transformation due to Blöchl [Phys. Rev. B 50, 17953 (1994)] to each calculated pseudo-orbital to obtain a corresponding frozen-core all-electron molecular orbital. Our model

Balázs Hetényi; Filippo de Angelis; Paolo Giannozzi; Roberto Car

2001-01-01

404

Deep Sea Coring  

NSDL National Science Digital Library

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.

Woods Hole Oceanographic Institution; Ocean and Climate Change Institute

405

Utah's New Mathematics Core  

ERIC Educational Resources Information Center

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…

Utah State Office of Education, 2011

2011-01-01

406

Coring the Ocean Floor  

NSDL National Science Digital Library

This site explains how core samples are taken from the ocean floor. Topics include how research cruises are planned, who makes up the crew of a research vessel, and what a cruise track is. Links to additional information are embedded in the text.

407

Theory of core excitons  

SciTech Connect

The observation of core excitons with binding energies much larger than those of the valence excitons in the same material has posed a long-standing theoretical problem. A proposed solution to this problem is presented, and Frenkel excitons and Wannier excitons are shown to coexist naturally in a single material. (GHT)

Dow, J. D.; Hjalmarson, H. P.; Sankey, O. F.; Allen, R. E.; Buettner, H.

1980-01-01

408

Physics of cluster cores  

Microsoft Academic Search

The hot intracluster medium (ICM) in the cores of most clusters of galaxies has a radiative cooling time of a few Gyr or less. XMM-Newton and Chandra data show however that the gas cools by at most a factor of three in temperature. This is the 'cooling flow problem', solutions to which will be discussed. Several involve the transport properties

A. C. Fabian

2004-01-01

409

Core, Canon, Curriculum.  

ERIC Educational Resources Information Center

Noting that higher education across the centuries has constituted a continuing dialogue between the minds of ancestors and of contemporaries, this paper traces the history of the common or core curriculum at the university level and warns against the current state of affairs. The paper proposes that a pedagogy is needed that can both discern and…

Levin, Harry

410

Navagating the Common Core  

ERIC Educational Resources Information Center

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…

McShane, Michael Q.

2014-01-01

411

Learning Core Meanings.  

ERIC Educational Resources Information Center

An interactive vocabulary learning technique is described that uses a "core meaning" approach to help solve some of the problems learners have with English homonyms. The technique was used successfully with a class of adult learners but can be adapted for younger and less proficient learners. (four references) (LB)

Visser, Annette

1989-01-01

412

Nucleosome Core Particle  

NASA Technical Reports Server (NTRS)

Nucleosome Core Particle grown on STS-81. The fundamental structural unit of chromatin and is the basis for organization within the genome by compaction of DNA within the nucleus of the cell and by making selected regions of chromosomes available for transcription and replication. Principal Investigator's are Dr. Dan Carter and Dr. Gerard Bunick of New Century Pharmaceuticals.

1997-01-01

413

Renewing the Core Curriculum  

ERIC Educational Resources Information Center

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…

Lawson, Hal A.

2007-01-01

414

The Earth's Core.  

ERIC Educational Resources Information Center

The nature of the earth's core is described. Indirect evidence (such as that determined from seismological data) indicates that it is an iron alloy, solid toward its center but otherwise liquid. Evidence also suggests that it is the turbulent flow of the liquid that generates the earth's magnetic field. (JN)

Jeanloz, Raymond

1983-01-01

415

Electromagnetic pump stator core  

DOEpatents

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.

Fanning, Alan W. (San Jose, CA); Olich, Eugene E. (Aptos, CA); Dahl, Leslie R. (Livermore, CA)

1995-01-01

416

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...

417

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....

418

Ultrasonic Drilling and Coring  

NASA Technical Reports Server (NTRS)

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.

Bar-Cohen, Yoseph

1998-01-01

419

Some Core Contested Concepts  

ERIC Educational Resources Information Center

Core concepts of language are highly contested. In some cases this is legitimate: real empirical and conceptual issues arise. In other cas