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1

Verification of LOGOS nodal method with heterogeneous burnup calculations for a BWR core  

Microsoft Academic Search

In the nodal method, node-homogenized constants are generally prepared using single assembly depletion calculations. In an actual core with mixed fuel loading, however, spectral interactions between assemblies cause neutron movements across the node surfaces. We must consider changes to the node-homogenized constants caused by instantaneous spectral interactions, as well as historical effects due to assembly depletion with spectral interactions. We

T. Iwamoto; M. Yamamoto; M. Tsuiki

1994-01-01

2

Burn-up reactivity measurements of the Joyo MK-II core.  

National Technical Information Service (NTIS)

The core averaged burn-up reactivity has been measured and calculated for the Joyo MK-II core. In order to evaluate the relationship between the calculational error of burn-up reactivity and the nuclear data or calculated neutron flux, the burn-up reactiv...

A. Yoshida Y. Nagaoki

1997-01-01

3

Power excursion analysis for high burnup cores  

SciTech Connect

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

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

1996-02-01

4

Burnup calculation methodology in the serpent 2 Monte Carlo code  

SciTech Connect

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

Leppaenen, J. [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland); Isotalo, A. [Aalto Univ., Dept. of Applied Physics, P.O.Box 14100, FI-00076 AALTO (Finland)

2012-07-01

5

Calculation of burnup of a black neutron absorber  

SciTech Connect

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

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

2011-12-15

6

Core Burnup Characteristics of High Conversion Light Water Reactor, (1). Core Analyses for HCLWR-J1 (V/Sub M//V/sub p/ Approx. =0.8).  

National Technical Information Service (NTIS)

In order to evaluate core burnup characteristics of a high conversion light water reactor (HCLWR) with tight pitched lattice, core burnup calculation was made using two dimensional diffusion method. The volume ratio of moderator to fuel is about 0.8 in th...

K. Okumura Y. Ishiguro E. Doi

1988-01-01

7

Analyses of Burnup at Plutonium Spots in Uranium-Plutonium Mixed Oxide Fuels in Light Water Reactors by Neutron Transport and Burnup Calculations  

Microsoft Academic Search

Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs

Takanori KAMEYAMA; Akihiro SASAHARA; Tetsuo MATSUMURA

1997-01-01

8

Reactor critical benchmark calculations for burnup credit applications  

SciTech Connect

In the criticality safety analyses for the development and certification of spent fuel casks, the current approach requires the assumption of fresh fuel'' isotopics. It has been shown that the removal of the fresh fuel'' assumption and the use of spent fuel isotopics ( burnup credit'') greatly increases the payload of spent fuel casks by reducing the reactivity of the fuel. Regulatory approval of burnup credit and the requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. Criticality analyses for low-enriched lattices of fuel pins using the fresh fuel isotopics'' assumption have been widely benchmarked against applicable critical experiments. However, the same computational methods have not been benchmarked against criticals containing spent fuel because of the non-existence of spent fuel critical experiments. Commercial reactors offer an excellent and inexhaustible source of critical configurations against which criticality analyses can be benchmarked for spent fuel configurations. This document provides brief descriptions of the benchmarks and the computational methods for the criticality analyses. 8 refs., 1 fig., 1 tab.

Renier, J.P.; Parks, C.V.

1990-01-01

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

Using Laguerre polynomials to compute the matrix exponential in burnup calculations  

SciTech Connect

An essential part of burnup analysis is to solve the burnup equations. The burnup equations can be regarded as a first-order linear system and solved by means of matrix exponential methods. Because of its large spectrum, it is difficult to compute the exponential of the burnup matrix. Conventional methods of computing the matrix exponential, such as the truncated Taylor expansion and the Pade approximation, are not applicable to burnup calculations. Recently the Chebyshev Rational Approximation Method (CRAM) has been applied to solve burnup matrix exponential and shown to be robust and accurate. However, the main defect of CRAM is that its coefficients are not easy to obtain. In this paper, an orthogonal polynomial expansion method, called Laguerre Polynomial Approximation Method (LPAM), is proposed to compute the matrix exponential in burnup calculations. The polynomial sequence of LPAM can be easily computed in any order and thus LPAM is quite convenient to be utilized into burnup codes. Two typical test cases with the decay and cross-section data taken from the standard ORIGEN 2.1 libraries are calculated for validation, against the reference results provided by CRAM of 14 order. Numerical results show that, LPAM is sufficiently accurate for burnup calculations. The influences of the parameters on the convergence of LPAM are also discussed. (authors)

She, D.; Zhu, A.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ., Beijing, 100084 (China)

2012-07-01

12

Reconstruction of pin power and burnup characteristics from nodal calculations in hexagonal-Z geometry  

Microsoft Academic Search

This paper reports on a reconstruction method which is developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D\\/REBUS-3 code system. Intranodal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the

W. S. Yang; P. J. Finck; H. Khalil

1992-01-01

13

Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry  

Microsoft Academic Search

A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D\\/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested

W. S. Yang; P. J. Finck; H. S. Khalil

1990-01-01

14

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

Microsoft Academic Search

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

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

1984-01-01

15

Criticality reference benchmark calculations for burnup credit using spent fuel isotopics  

SciTech Connect

To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs.

Bowman, S.M.

1991-04-01

16

Reconstruction of pin power and burnup characteristics from nodal calculations in hexagonal-Z geometry  

SciTech Connect

This paper reports on a reconstruction method which is developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intranodal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method is tested by performing several fast reactor numerical benchmark calculations and by comparing predicted local burnups with values measured for experimental assemblies in the experimental Breeder Reactor II. The results indicate that the reconstruction methods are quite accurate yielding maximum errors in power and nuclide densities that are {lt}2% for driver assemblies and typically {lt}5% for blanket assemblies.

Yang, W.S.; Finck, P.J.; Khalil, H. (Argonne National Lab., Argonne, IL (US))

1992-05-01

17

Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry  

SciTech Connect

A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested by performing several numerical benchmark calculations and by comparing predicted local burnups to values measured for experimental assemblies in EBR-11. The results indicate that the reconstruction methods are quite accurate, yielding maximum errors in power and nuclide densities that are less than 2% for driver assemblies and typically less than 5% for blanket assemblies. 14 refs., 2 figs., 5 tabs.

Yang, W.S.; Finck, P.J.; Khalil, H.S.

1990-01-01

18

Burnup calculation of fusion–fission hybrid energy system with thorium cycle  

Microsoft Academic Search

A fusion–fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In our group, a calculation system for analysis of fusion–fission hybrid reactor has been developed and various transport and burnup calculations were carried out for hybrid systems with U–Pu fuel cycle and ITER model so far. It was confirmed that such system is feasible

M. Matsunaka; S. Shido; K. Kondo; H. Miyamaru; I. Murata

2007-01-01

19

Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations  

Microsoft Academic Search

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity\\/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The

Nuria García-Herranz; Oscar Cabellos; Javier Sanz; Jesús Juan; Jim C. Kuijper

2008-01-01

20

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

21

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

SciTech Connect

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

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

2006-07-01

22

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations  

SciTech Connect

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yusung-gu, Taejon (Korea, Republic of)

2005-05-24

23

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations  

NASA Astrophysics Data System (ADS)

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

2005-05-01

24

Using ORIGEN and MCNP to calculate reactor criticals and burnup effects  

SciTech Connect

The purpose of this modeling effort was to verify the applicability of using ORIGEN-S and MCNP to the analysis of spent fuel of various enrichments and burnups. By comparing the results of criticality studies using MCNP and ORIGEN-S with the measured k{sub eff} of 1.0, the suitability of the coupled ORIGEN-S/ MCNP package was determined. This study presents the results of the benchmark modeling of five pressurized water reactor (PWR) critical configurations. For these analyses, a combination of ORIGEN-S and MCNP was used to analyze the fuel depletion and criticality of five power reactor core configuration.

Bowen, D.; Busch, R.D. [Univ. of New Mexico, Albuquerque, NM (United States)

1997-12-01

25

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

NASA Astrophysics Data System (ADS)

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.

Sterbentz, James W.

2009-08-01

26

Startup of “Candle” burnup in fast reactor from enriched uranium core  

Microsoft Academic Search

A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burnup reactivity control mechanisms such as control rods are not required.This burnup strategy can derive many

Hiroshi Sekimoto; Seiichi Miyashita

2006-01-01

27

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

28

Benchmark calculation of JMTRC core with MCNP  

SciTech Connect

It has not been verified that MCNP can be used in complicated core calculations, with plate-type fuel such as the JMTR. The JMTRC is a critical facility, and simulates the JMTR core, but the core is clean in comparison to the JMTR. The applicability of MCNP to core calculation with plate-type fuel was studied using the JMTRC for the benchmark calculation.

Nagao, Yoshiharu; Shimakawa, Satoshi; Komori, Yoshihiro [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

1995-12-31

29

Burnup calculations and chemical analysis of irradiated fuel samples studied in LWR-PROTEUS phase II  

SciTech Connect

The isotopic compositions of 5 UO{sub 2} samples irradiated in a Swiss PWR power plant, which were investigated in the LWR-PROTEUS Phase II programme, were calculated using the CASMO-4 and BOXER assembly codes. The burnups of the samples range from 50 to 90 MWd/kg. The results for a large number of actinide and fission product nuclides were compared to those of chemical analyses performed using a combination of chromatographic separation and mass spectrometry. A good agreement of calculated and measured concentrations is found for many of the nuclides investigated with both codes. The concentrations of the Pu isotopes are mostly predicted within {+-}10%, the two codes giving quite different results, except for {sup 242}Pu. Relatively significant deviations are found for some isotopes of Cs and Sm, and large discrepancies are observed for Eu and Gd. The overall quality of the predictions by the two codes is comparable, and the deviations from the experimental data do not generally increase with burnup. (authors)

Grimm, P.; Guenther-Leopold, I. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Berger, H. D. [AREVA NP GmbH, FEEP, Bunsenstrasse 43, D-91058 Erlangen (Germany)

2006-07-01

30

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

31

Experimental results and analysis of core physics experiments, FUBILA, for high burn-up BWR full MOX cores  

SciTech Connect

JNES has been performing MOX core physics experiments, FUBILA, in the EOLE critical facility of the CEA Cadarache center with collaboration of a French Consortium (CEA and COGEMA). The experiments have been designed to obtain the core physics data of operating conditions of full MOX BWR cores consisting of high burn up BWR MOX assemblies. The experiments consisting of seven different core configurations started from January 2005 and will be completed by August 2006. Theoretical analysis of the experimental data has been also carried out using a deterministic code, SRAC, and a continuous energy Monte Carlo calculation code, MVP, with major nuclear data libraries, JENDL-3.3, 3.2, ENDF/B-VI and JEFF-3.1 for the first critical core. (authors)

Yamamoto, T.; Kikuchi, S.; Kawashima, K.; Kamimura, K. [Japan Nuclear Energy Safety Organization, 3-17-1, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

2006-07-01

32

Technique for sensitivity analysis of space- and energy-dependent burn-up calculations  

Microsoft Academic Search

A practical method is presented for sensitivity analysis of the very complex, space-energy dependent burn-up equations, in which the neutron and nuclide fields are coupled nonlinearly. The adjoint burn-up equations that are given are in a form which can be directly implemented into multi-dimensional depletion codes, such as VENTURE\\/BURNER. The data sensitivity coefficients can be used to determine the effect

M. L. Williams; J. R. White

1979-01-01

33

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

34

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

SciTech Connect

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is {approx}375 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the electrons and the produced neutrons and photons but the current version of MCNPX doesn't support depletion/burnup calculation of the subcritical system with the generated neutron source from the target. MCB can perform neutron transport and burnup calculation for subcritical system using external neutron source, however it cannot perform electron transport calculations. To solve this problem, a hybrid procedure is developed by coupling these two computer codes. The user tally subroutine of MCNPX is developed and utilized to record the information of the each generated neutron from the photonuclear reactions resulted from the electron beam interactions. MCB reads the recorded information of each generated neutron thorough the user source subroutine. In this way, the neutron source generated by electron reactions could be utilized in MCB calculations, without the need for MCB to transport the electrons. Using the source subroutines, MCB could get the external neutron source, which is prepared by MCNPX, and perform depletion calculation for the driven subcritical facility.

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

2009-06-09

35

Appropriate burnup measurements for transportation burnup credit.  

National Technical Information Service (NTIS)

This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measur...

D. Lancaster E. Fuentes

1997-01-01

36

Analysis of BWR high burnup fuel in LOCA conditions  

SciTech Connect

High Burnup Fuel Behaviour has been growing in importance since middle 80's when pellet microstructure changes (rim effect) and cladding oxidation rates increase were observed. Later on, Cadarache reactivity tests revealed cladding integrity failures below safety limits. These phenomena, occurred at high burnup, stressed the necessity of having a wide experimental data base that would allow to dispose non-extrapolated data of material properties submitted to higher burnups than 40000 MWd/TM and data of new materials at the same time. One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MTU rod average burnup. The technical bases to support those high burnup levels are being developed. One of the licensing points of concern is the behaviour of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/TM and H.B. Robinson at 67 GWd/MTU. When the ANL tests have been finished, a conservative Peak Cladding Temperature/ Equivalent Cladding Reacted (PCT/ECR) limit will be determine from the residual ductility tests to be applied to the high burnup fuel. This makes necessary to determine the behaviour of the high burnup fuel in LOCA conditions and to determine the available safety margin. In licensing LOCA calculations, corresponding to present core designs and future core designs, the calculated PCT and ECR values as a function of the fuel burnup could be used to determine the relative severity of LOCA for the high burnup fuel. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations. (authors)

Garcia Sedano, Pablo; Dey Navarro, Jose Manuel; Gallego Cabezon, Ines; Orive Moreno, Raul [IBERDROLA Ingenieria y Consultoria S.A.U., Avda. de Burgos, 8B, 28036 Madrid (Spain)

2004-07-01

37

Development and validation of a fast reactor core burnup code – FARCOB  

Microsoft Academic Search

A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under

P. Mohanakrishnan

2008-01-01

38

FRAPCON-3: A Computer Code for the Calculation of Steady-State, Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup.  

National Technical Information Service (NTIS)

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 ro...

G. A. Berna C. E. Beyer K. L. Davis D. D. Lanning

1997-01-01

39

Finite element method for ferrite core loss calculation  

Microsoft Academic Search

A calculation method for ferrite core loss by finite element analysis (FEA) is developed based on core loss measurements for sample toroidal cores. Material properties extracted from these tests are then used in an FEA solver to predict core loss in a range of other core shapes. Measured data is used to evaluate the accuracy of the FEA simulations

Ping Han; Glenn R. Skutt; Ju Zhang; Fred C. Lee

1995-01-01

40

Comparison of the Burnup Characteristics and Radiotoxicity Hazards of Rock-like Oxide Fuel with Different Types of Additives  

Microsoft Academic Search

Burnup characteristics of rock-like oxide (PuO2-ZrO2: ROX) fueled LWR have been studied with an additive ThO2, UO2 or Er2O3. A ROX fueled LWR core has reactivity coefficients and burnup reactivity swing problems, which can be mitigated by these additives. From the burnup calculations of the ROX fueled LWR, it was found that the additive ThO2 causes a few percent decrease

Afroza SHELLEY; Hiroshi AKIE; Hideki TAKANO; Hiroshi SEKIMOTO

2001-01-01

41

Use of the local Pu239 concentration as an indicator of burnup spectral history in DYN3D  

Microsoft Academic Search

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd\\/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors

I. Bilodid; S. Mittag

2010-01-01

42

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

43

ULOF and UTOP Analyses of a Large Metal Fuel FBR Core Using a Detailed Calculation System  

Microsoft Academic Search

ULOF and UTOP analyses of a large metal fuel FBR core (1,500 MWe, averaged discharge burnup: 150 GWd\\/t) are conducted. The effect of core radial expansion is considered as the major negative feedback during the transient. A detailed analysis system is used, in which a transient core thermal-hydraulic code is coupled with three dimensional core radial deformation and reactivity feedback

Takeshi YOKOO; Hirokazu OHTA

2001-01-01

44

Application of Improved Coarse Mesh Method to BWR Core Calculations  

Microsoft Academic Search

The improved coarse mesh method, which was originally derived by Askew and applied to fast reactor core calculations by one of the authors, has been applied to boiling water reactor core calculations. Since direct application of the method leads to negative fluxes in the reflector region, a procedure has been developed in which the correction factor for coarse meshes is

Toshikazu TAKEDA; Etsuro SAJI

1980-01-01

45

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

46

Uncertainty in the power distribution for a fast reactor burnup cycle  

SciTech Connect

Demonstration that advanced reactor designs satisfy safety and performance goals requires the analysis of uncertainties in calculated reactor characteristics. Two of the important performance characteristics of advanced liquid metal reactors (LMR) are the burnup reactivity swing and the local power peaking factor. Previous work reported a study of the uncertainty in the burnup reactivity swing attributable to nuclear data uncertainties for a typical advanced LMR design. The results are reported here of some recent investigations into the uncertainty in the core power distribution after the completion of a burnup cycle. 11 refs., 1 fig., 2 tabs.

Downar, D.J.; Khalil, H. (Purdue Univ., Lafayette, IN (USA); Argonne National Lab., IL (USA))

1989-01-01

47

The impact of ENDF/B-V fission spectrum parameter uncertainty on the prediction of fast reactor burnup performance  

SciTech Connect

The sensitivity and uncertainty of various core burnup performance quantities (e.g., k[sub eff], burnup reactivity swing, local power density, etc.) to the heavy isotope fission spectrum parameters was investigated using depletion perturbation methods and ENDF/B-V covariance data. A brief description of the methods is followed by results of a 900-MW(thermal) fast reactor. The analysis here indicates that for a 900-MW(thermal) heterogeneous fast reactor, the experimental error in the ENDF/B-V fission spectrum parameters does not have a major impact on calculational uncertainty in the core burnup performance.

Downar, T.J.; Broda, J.; Kritzer, J. (Purdue Univ., Lafayette, IN (United States))

1990-01-01

48

Kinetics Calculations for RIA Experiments in the Capsule Driver Core.  

National Technical Information Service (NTIS)

Space-time kinetics calculations have been done for transient fuel behavior tests that were conducted during the late 1960's in the Capsule Driver Core in PBF. The purpose of the calculations was to determine the amount of energy deposited by delayed-neut...

A. J. Scott D. W. Nigg J. L. Judd S. A. Easson

1981-01-01

49

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

50

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -  

Microsoft Academic Search

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO2 fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (<

Tetsuo Matsumura; Takanori Kameyama; Yasushi Nauchi; Izumi Kinoshita

2005-01-01

51

Flux distribution and core loss calculation for single phase and five limb three phase transformer core designs  

Microsoft Academic Search

This paper presents results of an analytical core loss calculation method developed for single phase and five limb three phase core types. Shown are calculation results of flux distributions, flux density waveshapes, spatial flux density curves and loss distributions in cores other than the 3 phase-3 limb core. The analytical results show that the flux distribution is essentially uniform in

E. G. teNyenhuis; G. F. Mechler; R. S. Girgis

2000-01-01

52

BURNUP OF OMR FUEL ELEMENTS, HB1 AND HT1  

Microsoft Academic Search

Preliminary results from Battelle, on the first OMR Fuel Elements, ; indicate that the maximum burnup was approximately 0.22 at.% U for the high ; burnup element and 0.20 at.% U for the heat transfer element. In both cases, the ; maximum burnup occurred at the edge of the fuel plate facing the center of the ; reactor core at

1959-01-01

53

Recent Developments in No-Core Shell-Model Calculations  

Microsoft Academic Search

We present an overview of recent results and developments of the no-core shell model (NCSM), an ab initio approach to the nuclear many-body problem for light nuclei. In this aproach, we start from realistic two-nucleon or two- plus three-nucleon interactions. Many-body calculations are performed using a finite harmonic-oscillator (HO) basis. To facilitate convergence for realistic inter-nucleon interactions that generate strong

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

2009-01-01

54

Bethe-Salpeter equation calculations of core excitation spectra  

NASA Astrophysics Data System (ADS)

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

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

2011-03-01

55

PWR AXIAL BURNUP PROFILE ANALYSIS  

SciTech Connect

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

J.M. Acaglione

2003-09-17

56

No-core full configuration calculations for neutron droplets  

NASA Astrophysics Data System (ADS)

We present recent results from no-core full configuration calculations for neutron droplets in an external field. We use several different NN interactions (JISP16, chiral N3LO, and Minnesota) and discuss the similarities and differences in the obtained energies, rms radii, and density profiles. We get good numerical convergence for up to 20 neutrons, and our results are in agreement with other methods. These results form an excellent basis for validation and verification of new energy-density functionals for nuclear physics.

Vary, James; Maris, Pieter

2010-11-01

57

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

58

Ftr Tag Burnup.  

National Technical Information Service (NTIS)

The gas tag burnup changes investigated were limited to the three tags (Kr-78/Kr-80, Xe-126/Xe-129 and Kr-82/Kr-80) currently accepted as being the most desirable. Control rod tag burnup was significantly greater than fuel rod tag burnup. This occurs beca...

R. B. Kidman

1976-01-01

59

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

60

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

61

Accident Source Terms for Pressurized Water Reactors with High-Burnup Cores Calculated Using MELCOR 1.8.5.  

National Technical Information Service (NTIS)

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product rele...

D. A. Powers M. T. Leonard P. Longmire R. O. Gauntt S. G. Ashbaugh

2010-01-01

62

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

Microsoft Academic Search

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the

Randall O. Gauntt; Dana Auburn Powers; Scott G. Ashbaugh; Mark Thomas Leonard; Pamela Longmire

2010-01-01

63

Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses.  

National Technical Information Service (NTIS)

This report presents studies performed to support the development of a technical justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup...

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

2003-01-01

64

Conservative axial burnup distributions for actinide-only burnup credit.  

National Technical Information Service (NTIS)

Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup...

C. Kang D. Lancaster

1997-01-01

65

Economic benefits of increased discharge burnup for PWR. Prospect of increasing discharge burnup for Daya Bay NPP and Qinshan NPP1.  

National Technical Information Service (NTIS)

Increased discharge burnup brings a great deal of benefits to the utilities. The total fuel cost and its fraction in different fuel cycle activities have been calculated at different discharge burnup level and given specific conditions by using DQUECO cod...

D. Liu

1991-01-01

66

Hartree-Fock calculations of crystalline structure of nuclear matter with hard-core potentials  

Microsoft Academic Search

A Hartree-Fock calculation for the binding energy of a crystalline structure of nuclear matter in the presence of a hard-core potential is performed. As expected, a strongly unbound nuclear matter is obtained. NUCLEAR STRUCTURE Nuclear matter, Hartree-Fock calculations, hard-core potential.

B. Grammaticos; D. Vautherin

1979-01-01

67

Core Modification to Improve Irradiation Efficiency of the Experimental Fast Reactor Joyo  

Microsoft Academic Search

Core modification has been investigated to further increase the core burnup and to improve the irradiation efficiency of the experimental fast reactor Joyo. This modification enables the core to accommodate more irradiation test subassemblies that have lower fissile material contents compared with the driver fuel. The design calculations showed that the replacement of the radial reflector elements made of stainless

Shigetaka MAEDA; Masaya YAMAMOTO; Tomonori SOGA; Takashi SEKINE; Takafumi AOYAMA

2011-01-01

68

In-core thermal-hydraulic and fission product calculations for severe fuel damage analyses  

SciTech Connect

Best-estimate calculations of realistic fission product source terms are presented for the Severe Fuel Damage (SFD) tests conducted in the Power Burst Facility (PBF), utilizing the Advanced Reactor Severe Accident Program (ARSAP) bulk mass transfer correlation. Computer codes were written to perform the thermal-hydraulic and fission product calculations for the SFD tests. Fewer and slower releases are predicted with the ARSAP mass transfer correlation, in good agreement with the test results. The ARSAP mass transfer model correlates the inverse fuel temperature with the product of release rate and grain size considering the fuel/cladding interaction. The empirical coefficients were developed from Oak Ridge National Laboratory (ORNL) high-burnup fuel data in the 770 to 2,275 K temperature range. The ORNL test data indicate that the fuel/cladding interaction takes effect above 2,000 K.

Suh, K.Y.; Sharon, A.; Hammersley, R.J. (Fauske Associates, Inc., Burr Ridge, IL (USA))

1989-11-01

69

Accounting for the Breit interaction in relativistic effective core potential calculations of actinides  

NASA Astrophysics Data System (ADS)

The accuracy of accounting for the Breit effects in calculations of atoms and molecules in the framework of the two-component relativistic effective core potential (RECP) formalism is analysed theoretically and illustrated in calculations. Contributions of the Breit interaction between different core and valence shells are studied in the framework of the four-component SCF approximation for uranium and plutonium. It is shown that two-electron Breit effects between valence electrons can be neglected for the calculation of valence energies in systems containing actinides with 'chemical accuracy' (1 kcal mol-1 or 350 cm-1), whereas large core-core and core-valence Breit contributions can be efficiently described by one-electron RECP operators. Different versions of generalized RECPs with the Breit interaction taken into account are constructed for uranium and plutonium and compared for accuracy with the corresponding all-electron four-component calculations.

Petrov, A. N.; Mosyagin, N. S.; Titov, A. V.; Tupitsyn, I. I.

2004-12-01

70

Quantum confinement effect in Si\\/Ge core-shell nanowires: First-principles calculations  

Microsoft Academic Search

The electronic structure of Si\\/Ge core-shell nanowires along the [110] and [111] directions are studied with first-principles calculations. We identify the near-gap electronic states that are spatially separated within the core or the shell region, making it possible for a dopant to generate carriers in a different region. The confinement energies of these core and shell states provide an operational

Li Yang; Ryza N. Musin; Xiao-Qian Wang; M. Y. Chou

2008-01-01

71

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

72

Band structure calculations of Ge-Si core-shell nanowires  

Microsoft Academic Search

We have modeled and calculated the band structure of Ge-Si core-shell nanowire, with both subband interactions between Ge core and Si shell and the inhomogeneous strain effects taken into account. Our results show that the effective masses of subbands, the densities of states and quantum conductance will undergo significant changes and converge to saturated values, as the Si shell turns

Yuhui He; Chun Fan; Yu Ning Zhao; Jinfeng Kang; Xiao Yan Liu; Ruqi Han

2008-01-01

73

The Calculation of Contribution Rate of Education to the Regional Core Competence  

Microsoft Academic Search

Based on the production function of C-D, the calculation model measuring the contribution rate of education to enhance the core competence of a region was established, and the determination method of the parameters in the model was also defined. At last the contribution rate of different type of education in Jiangsu Province to the core competence of the region was

Hongzhuan Chen; Qiangqiang Zhao

2008-01-01

74

Full vectorial finite element method for acoustic mode calculation of suspended core fiber  

Microsoft Academic Search

Characterization of acoustic modes is important in the study of Brillouin gain spectrum. In this paper, acoustic modes of a suspended core fiber are numerically calculated by finite element method. Due to high contrast between mechanical coefficient of silica and air, full vectorial finite element method is used. Since core of SCF is surrounded by large air holes, the solutions

Soodabeh Nouri Jouybari; Hamid Latifi; Fahimeh Salehpoor; Faramarz Farahi

2011-01-01

75

Design of saturated iron core superconducting fault current limiter based on numerical calculation  

Microsoft Academic Search

The saturated iron core superconducting fault current limiter (SCFCL) utilizes the non-linear of ferromagnetic hysteresis to change its inductance. Its cores are saturated deeply under normal operation but driven out of saturation whenever fault occurs, so it is very difficult to design this SCFCL through theoretical calculation and there will be biggish error if deal the magnetization curve with subsection

Zhang Xuhong; Zhou Youqing; Zhang Zhifeng

2005-01-01

76

Molecular Evolution of A First Core in 3 Dimensional Hydrodynamic Calculations  

NASA Astrophysics Data System (ADS)

It is well established that stars are formed by gravitational collapse of molecular cloud cores. Collapsing cores initially undergo isothermal collapse. The isothermal condition breaks down at the density of ˜ 10-13 g cm-3, and the temperature starts rising. Increasing gas pressure decelerates the contraction, and the cores come to hydrostatic equilibrium with a radius of a few AU and a mass of ˜ 0.01 M?, which is called the first cores (e.g. Larson 1969). Observation of the first cores is important but challenging, since their lifetime is short (˜ 1000 yr). The mechanical property of the first cores have been studied by multi-dimensional hydrodynamic calculations considering interstellar magnetic fields and radiative transfer (e.g. Tomisaka 2002; Machida et al.2008; Tomida et al. 2010). In contrast, their chemical property is yet to be understood. It is important to reveal their chemical property in terms of which lines we should use to observe the first cores. In addition, the first cores evolve to protoplanetary disks (Saigo et al. 2008; Machida et al. 2010), hence the compositions of the first cores restrict the initial compositions of disks. We investigate molecular evolution of star forming cores that are initially rotating molecular cloud cores and collapse to form the first cores. The results of three dimensional hydrodynamic calculations (Matsumoto & Hanawa 2003) are adopted as physical models of the core. We trace trajectories of test particles in the hydrodynamic calculations, and molecular evolution is solved using low temperature chemical network (Garrod & Herbst 2006) at T < 100 K and high temperature network (Harada et al. 2010) at T > 100 K along the trajectories. We also consider three body reactions and collisional dissociations (Willacy et al. 1998). Trace particles fall into the first core almost spherically, and rotate in the first core where the spiral arms transports angular momentum. In our model with barotropic approximation, we find that in outer regions (R > 5 AU), the composition is similar to the low temperature chemistry. In intermediate regions (R ˜3 AU), hot-core like species, such as HCOOCH_3 and CH_3OCH_3 are generated. In central regions (R < 1 AU), complex molecules, such as HC_7N, HC_9N and NH_2CN, are formed in the gas phase.

Furuya, K.; Aikawa, Y.; Matsumoto, T.; Tomida, K.; Saigo, K.; Tomisaka, K.; Hersant, F.; Wakelam, V.

2011-05-01

77

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

78

Fuel element powers, STVU masses, and burnups from gamma-scanning data: Preliminary analysis of irradiated ORR (Oak Ridge Research Reactor) LEU fuel elements  

SciTech Connect

Fuel elements used in the ORR whole-core LEU fuel demonstration have been gamma-scanned to determine axial distributions of UZLa and TXCs fission product activities. This data has been analyzed to determine cycle-averaged fuel element powers, residual STVU masses, and burnups of discharged fuel elements. Methods used to analyze the data are discussed and results are presented for the LEU fuel elements. Measured and calculates fuel element powers agree to within 5%, residual STVU masses to within 2%, and burnups to within 3%. These results are somewhat preliminary and await improved burnup calculations and independent calibration data to be based on the destructive analyses of a number of irradiated fuel elements. 4 refs., 7 figs., 3 tabs.

Bretscher, M.M.; Snelgrove, J.L.; Hobbs, R.W.

1988-01-01

79

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

SciTech Connect

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

Sambuu, Odmaa; Nanzad, Norov [Nuclear Research Center National University of Mongolia Ulaanbaatar (Mongolia)

2009-03-31

80

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

81

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

82

Melting of iron under Earth's core conditions from diffusion Monte Carlo free energy calculations.  

PubMed

The temperature of Earth's core is a parameter of critical importance to model the thermal structure of Earth. Since the core is mainly made of iron, with a solid liquid boundary (the inner core boundary) at 1220 km from the center of the Earth, the melting temperature of iron at the pressure of the ICB provides constraints on the temperature of the core. These constraints are based either on extrapolations to ICB pressure of experimental measurements, or on theoretical calculations which employed various flavors of quantum mechanics, most notably density functional theory. Significant disagreement between estimates obtained with different methods calls for calculations based on more accurate techniques. Here we used quantum Monte Carlo techniques to compute the free energies of solid and liquid iron at ICB conditions. We obtained an iron melting temperature at 330 GPa of 6900+/-400 K. PMID:19792692

Sola, Ester; Alfè, Dario

2009-08-14

83

Performance of relativistic effective core potentials in DFT calculations on actinide compounds.  

PubMed

Density functional theory (DFT) calculations using relativistic effective core potentials (RECPs) have emerged as a robust and fast method of calculating the structural parameters and energy changes of the thermochemical reactions of actinide complexes. A comparative investigation of the performance of the Stuttgart small-core and large-core RECPs in DFT calculations has been carried out. The vibrational frequencies and reaction enthalpy changes of several uranium(VI) compounds computed using these RECPs were compared to those obtained using DFT and a four-component one-electron scalar relativistic approximation of the full Dirac equation with large all-electron basis sets (AE). The relativistic AE method is a full solution of the Dirac equation with all spin components separated out. This method gives the "correct" answer (with respect to scalar relativity) which should be closest to experimental values when an adequate density functional is used and in the absence of significant spin-orbit effects. The small-core RECP always show better agreement with the four-component scalar- relativistic AE method than the large-core RECP. We conclude that the 5s, 5p, and 5d orbitals are of great importance in determining the chemistry of actinide complexes. Instances in which large-core RECPs give better agreement with experimental data are attributed to either experimental uncertainties or error cancellations. PMID:20039716

Odoh, Samuel O; Schreckenbach, Georg

2010-02-01

84

Ab initio effective core potentials for molecular calculations. Potentials for K to Au including the outermost core orbitals  

SciTech Connect

Ab initio effective core potentials (ECP's) have been generated to replace the innermost core electron for third-row (K--Au), fourth-row (Rb--Ag), and fifth-row (Cs--Au) atoms. The outermost core orbitals: corresponding to the ns/sup 2/np/sup 6/ configuration for the three rows here: are not replaced by the ECP but are treated on an equal footing with the nd, (n+1)s and (n+1)p valence orbitals. These ECP's have been derived for use in molecular calculations where these outer core orbitals need to be treated explicitly rather than to be replaced by an ECP. The ECP's for the forth and fifth rows also incorporate the mass--velocity and Darwin relativistic effects into the potentials. Analytic fits to the potentials are presented for use in multicenter integral evaluation. Gaussian orbital valence basis sets are developed for the (3s, 3p, 3d, 4s, 4p), (4s, 4p, 4d, 5s, 5p), and (5s, 5p, 5d, 6s, 6p) ortibals of the three respective rows.

Hay, P.J.; Wadt, W.R.

1985-01-01

85

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

86

Benchmark Tests of Radiation Transport Computer Codes for Reactor Core and Shield Calculations  

Microsoft Academic Search

Aiming at providing test problems that may be used to verify an adequate performance of the current version of a neutron and ?-ray transport computer code used for reactor core or shield calculations, we summarize the input data and the calculated results for three benchmark problems.The 1st problem deals with a 1-dimensional small spherical reactor for use to test 1-dimensional

Takumi ASAOKA; Norio ASANO; Hisashi NAKAMURA; Hiroshi MIZUTA; Hiroshi CHICHIWA; Tadahiro OHNISHI; Shun-ichi MIYASAKA; Atsushi ZUKERAN; Tsuneo TSUTSUI; Toichiro FUJIMURA; Satoru KATSURAGI

1978-01-01

87

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

88

BEAGL-01, a Computer Code for Calculating Rapid LWR Core Transients.  

National Technical Information Service (NTIS)

BEAGL-01 (Brookhaven's and EPRI's Adaptation of the TWIGL code) is a computer program for calculating the conditions in a light water reactor (LWR) core at steady state and during transients. It solves the finite-difference neutron kinetics equations on a...

D. J. Diamond A. L. Aronson

1983-01-01

89

Overview of a burnup credit issues  

SciTech Connect

A traditional assumption used in evaluating the criticality safety of a spent fuel storage or transport cask is that the spent fuel is as reactive as fresh fuel. This known as the fresh fuel assumption.'' The assumption avoids a number of calculational and verification problems, but takes a heavy toll in decreased efficiency. An alternative to the fresh fuel assumption is called burnup credit.'' That is, the reduced reactivity of spent fuel that occurs from the net depletion of fissile nuclides and the net increase in fission and activation product neutron absorbers (poisons) is considered. Burnup credit has been successfully applied to spent fuel storage pools in the US, resulting in increased capacity and permitting the storage of spent fuel with higher initial enrichments. Both of the commercial transport cask designers supporting the Cask Systems Development Program (CSDP) of the US Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) have cask designs incorporating burnup credit. This document provides a discussion of burnup credit as applied to spent fuels.

Sanders, T.L.; Seager, K.D.; Ewing, R.I.

1991-01-01

90

Overview of a burnup credit issues  

SciTech Connect

A traditional assumption used in evaluating the criticality safety of a spent fuel storage or transport cask is that the spent fuel is as reactive as fresh fuel. This known as the ``fresh fuel assumption.`` The assumption avoids a number of calculational and verification problems, but takes a heavy toll in decreased efficiency. An alternative to the fresh fuel assumption is called ``burnup credit.`` That is, the reduced reactivity of spent fuel that occurs from the net depletion of fissile nuclides and the net increase in fission and activation product neutron absorbers (poisons) is considered. Burnup credit has been successfully applied to spent fuel storage pools in the US, resulting in increased capacity and permitting the storage of spent fuel with higher initial enrichments. Both of the commercial transport cask designers supporting the Cask Systems Development Program (CSDP) of the US Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM) have cask designs incorporating burnup credit. This document provides a discussion of burnup credit as applied to spent fuels.

Sanders, T.L.; Seager, K.D.; Ewing, R.I.

1991-12-31

91

Double-core-polarization contribution to atomic parity-nonconservation and electric-dipole-moment calculations  

NASA Astrophysics Data System (ADS)

We present a detailed study of the effect of double core polarization (the polarization of core electrons due to the simultaneous action of the electric dipole and parity-violating weak fields) for amplitudes of the ss and sd parity-nonconserving transitions in Rb, Cs, Ba+, La2+, Tl, Fr, Ra+, Ac2+, and Th3+ as well as electron electric-dipole-moment enhancement factors for the ground states of the above neutral atoms and Au. This effect is quite large and has the potential to resolve some disagreement between calculations in the literature. It also has significant consequences for the use of experimental data in the accuracy analysis.

Roberts, B. M.; Dzuba, V. A.; Flambaum, V. V.

2013-10-01

92

Validation Work to Support the Idaho National Engineering and Environmental Laboratory Calculational Burnup Methodology Using Shippingport Light Water Breeder Reactor (LWBR) Spent Fuel Assay Data  

SciTech Connect

Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

J. W. Sterbentz

1999-08-01

93

Extended step characteristic model for quarter-core gamma heating calculations  

SciTech Connect

Discrete ordinates codes are seldom used in lattice or core calculation, because of their limitation to simple geometries, which can be represented using an orthogonal mesh in a given coordinate system. Rough geometric approximations are often applied to obtain an estimate for a flux distribution. However, other methods, such as integral transport or Monte Carlo approaches, are generally more suited to irregular geometries. Each of these methods has its own weaknesses: integral transport methods are limited to problems in which the angular variation of the flux is isotropic or linearly anisotropic; Monte Carlo methods can be time consuming. The extended step characteristic (ESC) method has been developed to apply the discrete ordinates approximation to complicated geometries for which other methods provide less satisfactory solutions. The CENTAUR code has been developed to solve the two-dimensional transport equation using the ESC approach. This paper presents results of CENTAUR calculations for a quarter-core gamma redistribution problem for the Savannah River site (SRS) K reactor, under drained tank conditions following a postulated double-ended guillotine break loss-of-coolant accident. The calculations were used to confirm TWOTRAN calculations, which were based on a coarse approximation of the core geometry. A comparison of the results serves to demonstrate the capabilities and efficiency of the ESC approach.

DeHart, M.D.; Webb, R.L. (Westinghouse Savannah River Company, Aiken, SC (United States))

1993-01-01

94

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

95

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

Microsoft Academic Search

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

D. Grigoriadis; M. Varvayanni; N. Catsaros; E. Stakakis

2008-01-01

96

TOPICAL REVIEW: Recent developments in no-core shell-model calculations  

Microsoft Academic Search

We present an overview of recent results and developments of the no-core shell model (NCSM), an ab initio approach to the nuclear many-body problem for light nuclei. In this approach, we start from realistic two-nucleon or two- plus three-nucleon interactions. Many-body calculations are performed using a finite harmonic-oscillator (HO) basis. To facilitate convergence for realistic inter-nucleon interactions that generate strong

Petr Navrátil; Sofia Quaglioni; Ionel Stetcu; Bruce R. Barrett

2009-01-01

97

BEAGL-01: a computer code for calculating rapid LWR core transients. Volume 2. User's manual  

Microsoft Academic Search

BEAGL-01 is a computer program for calculating the conditions in an LWR core at steady state and while undergoing a transient. It couples a neutron kinetic model in cylindrical (r,z) geometry with a two-phase thermal-hydraulic model for multiple parallel channels. Appropriate applications include reactivity insertion accidents in both pressurized reactors and BWRs and transients in BWRs initiated by perturbations in

A. L. Aronson; D. J. Diamond

1984-01-01

98

Parametric calculation of pulse transformer with open magnetic cores based on magnetostatic-field theory  

NASA Astrophysics Data System (ADS)

An analytical method for calculating the magnetostatic field of a pulse transformer with open magnetic cores is put forward in this paper, and formulas for calculating inductances of a small aspect-ratio transformer are derived. In comparison to results calculated by finite element magnetostatic-field simulations, the calculated values of inductance of primary winding L1 and the inductance of secondary winding L2 have a relative error of about 5%, while the error of the coupling coefficient (k) is less than 2%. Meanwhile, the effect of current nonuniformity in the primary winding on magnetizing inductance is studied. According to the calculated results, this effect reduces the magnetizing inductance and the coupling coefficient of the transformer, and can lead to an overvoltage phenomenon on the secondary winding. A small aspect-ratio pulse transformer with open magnetic cores is developed, which has a small size of 250mm×150mm in length and diameter, respectively. Inductances of the transformer are measured. The measured results conform to the law obtained in this work. Tests of the pulsed transformer are carried out. Experimental results show that the transformer can export a high-voltage pulse with an amplitude of 310 kV and full width at half maximum of 1?s.

Yu, Bin-xiong; Liu, Jin-liang

2013-01-01

99

Recent development of Monte Carlo shell model and its application to no-core calculations  

NASA Astrophysics Data System (ADS)

One of the major challenges in nuclear theory is to reproduce and to predict nuclear structure from ab initio calculations with realistic nuclear forces. As the current limitation of direct diagonalization of Hamiltonian matrices by Lanczos iteration method is around the order of matrix dimensionality 1010 in shell-model calculations, it is difficult to access heavier nuclei beyond the p shell with sufficiently large basis spaces. It is possible to overcome this difficulty by utilizing efficient approximate methods to reproduce full ab initio solutions with good precision and quantified uncertainties. Following the major success of the Monte Carlo shell model (MCSM) with an assumed inert core in the sd- and pf-shell regions and also by recent developments in the MCSM algorithm, the no-core MCSM is expected to be one of the most powerful tools to meet these conditions. We have performed benchmark calculations in the p-shell region. Results of energies are compared with those in the full configuration interaction and no-core full configuration methods. These are found to be consistent with each other within quoted uncertainties when they could be quantified. We also compare and discuss the radial density of the helium-4 ground state extracted from the MCSM and FCI many-body wave functions.

Abe, T.; Maris, P.; Otsuka, T.; Shimizu, N.; Tsunoda, Y.; Utsuno, Y.; Vary, J. P.; Yoshida, T.

2013-08-01

100

Burnup credit validation of SCALE-4 using light-water-reactor criticals  

SciTech Connect

The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison with LWR core criticals. These benchmarks are relevant because they test a methodology`s ability to predict spent fuel isotopic and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. As part of the Burnup Credit Analysis Verification (BCAV) Task, the U.S. Department of Energy has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs). The initial analysis methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power`s Slurry and North Anna reactors. However, the analysis procedure was complex and included the calculation of lumped fission products. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by a comparison with three reactor critical configurations from Tennessee Valley Authority`s Sequoyah Unit 2 Cycle 3 and two from Virginia Power`s Slurry Unit 1 Cycle 2.

Bowman, S.M.; Hermann, O.W. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Albuquerque, NM (United States)

1993-03-01

101

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

102

Analysis of mixed oxide fuel irradiated in EBR-II: measured vs. predicted burnup  

Microsoft Academic Search

The calculation of burnup in mixed-oxide fuel pins irradiated in EBR-II is shown to agree with burnup measured by postirradiation radiochemical analysis. The mean percent deviation is 0.12% with a variance of 2.49%. This gives a high level of confidence that HIST, which uses 91% of the nominal EBR-II power, provides an accurate estimate of burnup.

L. D. Scott; D. S. Dutt; D. R. Wilson

1978-01-01

103

High-burnup oxide fuel in European fast reactors  

Microsoft Academic Search

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

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

1989-01-01

104

Review of Information for Spent Nuclear Fuel Burnup Confirmation.  

National Technical Information Service (NTIS)

The Interim Staff Guidance on burnup credit (ISG-8, revision 2) for pressurized-water-reactor spent nuclear fuel in storage and transport casks, issued in 2002 by the U.S. Nuclear Regulatory Commissions Spent Fuel Project Office, recommends an out-of-core...

B. B. Bevard C. V. Parks J. C. Wagner M. Aissa

2009-01-01

105

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

NASA Astrophysics Data System (ADS)

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

Guo, Xing; Wu, Zhensen; Linghu, Longxiang

2013-09-01

106

No Core CI calculations for light nuclei with chiral 2- and 3-body forces  

NASA Astrophysics Data System (ADS)

The atomic nucleus is a self-bound system of strongly interacting nucleons. In No-Core Configuration Interaction calculations, the nuclear wavefunction is expanded in Slater determinants of single-nucleon wavefunctions (Configurations), and the many-body Schrödinger equation becomes a large sparse matrix problem. The challenge is to reach numerical convergence to within quantified numerical uncertainties for physical observables using finite truncations of the infinite-dimensional basis space. We discuss strategies for constructing and solving the resulting large sparse matrices for a set of low-lying eigenvalues and eigenvectors on current multicore computer architectures. Several of these strategies have been implemented in the code MFDn, a hybrid MPI/OpenMP Fortran code for ab initio nuclear structure calculations that scales well to over 200,000 cores. We discuss how the similarity renormalization group can be used to improve the numerical convergence. We present results for excitation energies and other selected observables for 8Be and 12C using realistic 2- and 3-body forces obtained from chiral perturbation theory. Finally, we demonstrate that collective phenomena such as rotational band structures can emerge from these microscopic calculations.

Maris, Pieter; Metin Aktulga, H.; Binder, Sven; Calci, Angelo; Çatalyürek, Ümit V.; Langhammer, Joachim; Ng, Esmond; Saule, Erik; Roth, Robert; Vary, James P.; Yang, Chao

2013-08-01

107

Burnup credit validation of SCALE-4 using light water reactor criticals  

SciTech Connect

The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water-reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison to LWR core criticals. These are relevant benchmarks because they test a methodology's ability to predict spent fuel isotopics and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. The US Department of Energy Burnup Credit Program has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs) in order to validate an appropriate analysis methodology. The initial methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power's Surry and North Anna reactors. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by comparison to three reactor critical configurations from Tennessee Valley Authority's Sequoyah Unit 2 Cycle 3 and two from Virginia Power's Surry Unit 1 Cycle 2.

Bowman, S.M.; Hermann, O.W. (Oak Ridge National Lab., TN (United States)); Brady, M.C. (Sandia National Labs., Albuquerque, NM (United States))

1993-01-01

108

Ab initio effective core potentials for molecular calculations. Potentials for main group elements Na to Bi  

SciTech Connect

A consistent set of ab initio effective core potentials (ECP) has been generated for the main group elements from Na to Bi using the procedure originally developed by Kahn. The ECP's are derived from all-electron numerical Hartree--Fock atomic wave functions and fit to analytical representations for use in molecular calculations. For Rb to Bi the ECP's are generated from the relativistic Hartree--Fock atomic wave functions of Cowan which incorporate the Darwin and mass--velocity terms. Energy-optimized valence basis sets of (3s3p) primitive Gaussians are presented for use with the ECP's. Comparisons between all-electron and valence-electron ECP calculations are presented for NaF, NaCl, Cl/sub 2/, Cl/sub 2//sup -/, Br/sub 2/, Br/sub 2//sup -/, and Xe/sub 2//sup +/. The results show that the average errors introduced by the ECP's are generally only a few percent.

Wadt, W.R.; Hay, P.J.

1985-01-01

109

Ab initio no-core Gamow shell model calculations with realistic interactions  

NASA Astrophysics Data System (ADS)

No-core Gamow shell model (NCGSM) is applied to study selected well-bound and unbound states of helium isotopes. This model is formulated on the complex energy plane and, by using a complete Berggren ensemble, treats bound, resonant, and scattering states on equal footing. We use the density matrix renormalization group method to solve the many-body Schrödinger equation. To test the validity of our approach, we benchmarked the NCGSM results against Faddeev and Faddeev-Yakubovsky exact calculations for 3H and 4He nuclei. We also performed ab initio NCGSM calculations for the unstable nucleus 5He and determined the ground-state energy and decay width, starting from a realistic N3LO chiral interaction.

Papadimitriou, G.; Rotureau, J.; Michel, N.; P?oszajczak, M.; Barrett, B. R.

2013-10-01

110

Whole-core neutron transport calculations without fuel-coolant homogenization  

SciTech Connect

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

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

2000-02-10

111

ITC / BBQ model for coupled core-edge impurity transport calculations  

NASA Astrophysics Data System (ADS)

Consideration of multi-species impurity transport in advanced regimes introduces the need to discriminate between the screening properties of the scrape-off layer and the core accumulation properties. To model these cases a new radial impurity transport code (ITC), incorporating atomic data from both the ADAS system and HULLAC, has been developed and coupled to the BBQ 3-D Monte Carlo impurity transport code. The BBQ calculation treats plasma-wall interactions (impurity generation and/or recycling) and provides boundary conditions in the form of charge-resolved edge influxes to ITC. In turn, the ITC step provides the effluxes of impurities to the edge code. Background plasma data in the core and scrape-off layer are needed, and the solution is constrained by edge and core spectroscopy. The coupled steps must be iterated to reach a steady state solution. Since, in principle, self-sputtering can lead to impurity density runaway ('C bloom') the convergence of this iteration is not assured, on physical grounds. Previously, a prototype of this coupling was demonstrated using the MIST and SANCO radial codes and a converged iteration has been demonstrated. Results from the ITC / BBQ coupling model are presented for the case of long pulse shots using the Tore Supra Composants Internes et Limiteurs (CIEL) configuration. [1] J. Hogan, E. Tsitrone et al, , EPS 2003, St Peterburg

Thomas, P. R.; Hogan, J.; Guirlet, R.; Lowry, C.; Schunke, B.

2003-10-01

112

Application of continuous-energy Monte Carlo code as a cross-section generator of BWR core calculations  

Microsoft Academic Search

A continuous-energy Monte Carlo code is newly applied for the assembly calculations of actual BWR core analysis. Few-groups cross-sections and related constants (kinetic parameters) were generated by the continuous-energy Monte Carlo code MVP-BURN, and were tabulated for a core simulator. The commercial BWR, HAMAOKA-3 (1100MWe:BWR-5), was analyzed by a coupled neutronic-thermalhydraulic core simulator based on modified one-group diffusion theory using

Masayuki Tohjoh; Masato Watanabe; Akio Yamamoto

2005-01-01

113

TOPICAL REVIEW: Recent developments in no-core shell-model calculations  

NASA Astrophysics Data System (ADS)

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

Navrátil, Petr; Quaglioni, Sofia; Stetcu, Ionel; Barrett, Bruce R.

2009-08-01

114

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

SciTech Connect

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

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

2012-07-01

115

Large-scale Bethe-Salpeter equation calculations of core-level x-ray spectra  

NASA Astrophysics Data System (ADS)

Recently an approach has been developed for Bethe-Salpeter equation (BSE) calculations of core-level x-ray spectra, which is implemented in the ocean package footnotetext J. Vinson, E. L. Shirley, J. J. Rehr, and J. J. Kas, Phys. Rev. B 83, 115106 (2011); J. Vinson and J. J. Rehr, Phys. Rev. B (in press , 2012) which combines plane-wave, pseudopotential DFT electronic structure, PAW transition elements, GW self-energy corrections, and the NIST BSE solver. The method yields both dipole limited and finite momentum transfer spectra. Here we discuss several recent advances which yield a unified treatment of both extended states and atomic multiplet effects. In particular our approach now includes spin-dependent potentials and hole-dependent lifetimes, and gives an improved treatment of L2,3 edges, where contributions to spectral weight come from a mix of two distinct core holes. We have also extended the code interface to include pseudopotential wave functions from abinit, QuantumEspresso, or an interpolation based scheme, thus enabling large-scale calculations with unit cells in excess of 2000 å^3. Applications to water and ice structures are briefly discussed.

Rehr, J. J.; Vinson, J.; Gilmore, K.

2013-03-01

116

Calculational-experimental research models for a fast reactor with a heterogeneous core  

SciTech Connect

The physical characteristics of heterogeneous metallic oxide cores were experimentally studied by physical tests of the critical assemblies BFS-46 and BFS-46AZ, which simulate a reactor of the BN-1600 type, into the core of which a fuel assembly with metallic uranium is inserted. A calculational model for the critical assemblies being investigated, showing the zones and their dimensions, is presented. The critical assembly BFS-46AZ is a modification of the basic critical assembly BFS-46 which adds plutonium to the IBZ to simulate its accumulation during reactor operation. The BFS-46 and BFS-46AZ assemblies have identical dimensions for the IBZ and LEZ, and have different HEZ dimensions, necessary to ensure the criticality of each assembly. Plutonium with a /sup 240/Pu content equal to 3.8% is used in the LEZ. The critically parameters are calculated using one-dimensional and two-dimensional models in a 26-group diffusion approximation based on the BNAP-78 system of group constants.

Belov, S.P.; Bobrov, S.B.; Kazanskii, Yu.A.; Kuzin, E.N.; Matveev, V.I.; Novozhilov, A.I.; Chernyi, V.A.

1987-11-01

117

From Non-Hermitian Effective Operators to Large-Scale No-Core Shell Model Calculations for Light Nuclei.  

National Technical Information Service (NTIS)

No-core shell model (NCSM) calculations using ab initio effective interactions are very successful in reproducing experimental nuclear spectra. The main theoretical approach is the use of effective operators, which include correlations left out by the tru...

B. R. Barrett I. Stetcu J. P. Vary P. Navratil

2006-01-01

118

Parameter Survey for Burnup of High Conversion Light Water Reactor Lattice.  

National Technical Information Service (NTIS)

Burnup calculations were made on a lattice model for high conversion light water reactor (HCLWR) in order to assess the feasibility of HCLWR concept and to obtain reference data on considering its neutronic characteristics. In these calculations, lattice ...

H. Akie Y. Ishiguro M. Ido

1987-01-01

119

End effects in the criticality analysis of burnup credit casks  

SciTech Connect

A study to evaluate the effect of axially dependent burnup on k{sub eff} has been performed as part of an effort to qualify procedures to be used in establishing burnup credit in shipping cask design and certification. This study was performed using a generic 31-element modular cast-iron cask (wall thickness 33.1 cm) with a 1-cm-thick borated stainless-steel basket for reactivity control. Fuel isotopics used here are those of the 17 {times} 17 Westinghouse assemblies from the North Anna Unit 1 reactor. Virginia Power (VP) provided detailed spatial isotopics for the fuel assemblies in-core at beginning-of-cycle 5 (BOC-5) as generated from their PDQ analyses. Twenty-two axial planes were defined in the original VP data. The isotopics used in this study were for a 3.41 initial wt % {sup 235}U and an average burnup of 31.5 GWd/MTU.

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

1990-01-01

120

Addressing the Axial Burnup Distribution in PWR Burnup Credit Criticality Safety.  

National Technical Information Service (NTIS)

This paper summarizes efforts related to developing a technically justifiable approach for addressing the axial burnup distribution in PWR burnup-credit criticality safety analyses. The paper reviews available data on the axial variation in burnup and the...

J. C. Wagner

2001-01-01

121

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

NASA Astrophysics Data System (ADS)

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 rib guide fields; the effective-index method, 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 degree(s)C and 300 degree(s)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 were 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

1995-09-01

122

Ab initio calculations of the elasticity of iron and iron alloys at inner core conditions: evidence for a partially molten inner core?  

NASA Astrophysics Data System (ADS)

The nature of the stable phase of iron in the Earth's solid inner core is still highly controversial, with both experiments and seismology suggesting the occurrence of an uncharacterised phase transformation at core conditions. Theory is also undecided; although the hcp phase is thermodynamically stable, there is some possibility that the bcc phase could be stabilised at core pressures and temperatures by light elements such as silicon. Knowledge of the elastic properties of the candidate phases is essential if we are to understand core structure, composition and evolution. Ab initio finite temperature molecular dynamics simulations have been used to calculate the elasticity of hcp-Fe, bcc-Fe, FeS and FeSi at core conditions. The calculated compressional wave velocities are in excellent agreement with the most recent experimental data, following a Birch's Law type trend with density, which is almost independent of temperature. However, the calculated shear wave velocities of all phases studied are significantly higher than those inferred from seismology. This discrepancy can only be explained if the inner core is partially molten containing more than 8% liquid.

Vocadlo, L.

2006-12-01

123

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

Microsoft Academic Search

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

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

2005-01-01

124

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

125

PIUS core performance analysis  

SciTech Connect

A detailed evaluation of the fuel-burnup dependent power distribution and the scram reactivity for the PIUS reactor design has been performed. The analyses were carried out using the CPM lattice physics and NODE-P2 core neutronics/thermal-hydraulics codes, and are based on the information provided in the PIUS Preliminary Safety Information Document. Cycle depletion calculations were performed for a set of nine representative initial core loadings and the three-dimensional core power distributions were determined. These calculations indicate that the PIUS radial F{sub {Delta}h} and total F{sub Q} power peaking is stronger than that indicated by the PIUS reference-design values. The scram reactivity resulting from the injection of highly borated pool water was calculated for a series of time-dependent linear ramp and square-wave pool flows. The three-dimensional distribution of the borated pool water throughout the core was modeled and the spatial reactivity effects of the distributed boron were determined. For pool flows that increase as a linear ramp, the spatial reactivity effects of the distributed boron were very small. In this case, a constant core-average boron reactivity coefficient can be used to model the PIUS scram reactivity.

Carew, J.F.; Aronson, A.; Cokinos, D.M.; Prince, A.; Selcow, E.C.

1996-03-01

126

BEAGL-01, a computer code for calculating rapid LWR core transients  

SciTech Connect

BEAGL-01 (Brookhaven's and EPRI's Adaptation of the TWIGL code) is a computer program for calculating the conditions in a light water reactor (LWR) core at steady state and during transients. It solves the finite-difference neutron kinetics equations on an r,z (radial, axial) mesh, the thermal-hydraulic equations for the coolant in multiple parallel, i.e., one-dimensional, channels and the one-dimensional radial fuel rod heat conduction equations for pellet, gap and clad. The analyst provides time dependent boundary conditions and/or specifications for control rod movement in order to perturb the system from an initial steady state. The boundary conditions are the inlet flow rate and temperature and a single system pressure. The analyst also supplies a normalized inlet flow distribution across the core which does not vary with time. Control rod movement includes the center rod by itself, all banks of control rods, or some combination of these. The objective of this summary is to give capabilities and limitations.

Diamond, D.J.; Aronson, A.L.

1983-01-01

127

Ab Initio No-Core Shell Model Calculations Using Realistic Two- and Three-Body Interactions  

SciTech Connect

There has been significant progress in the ab initio approaches to the structure of light nuclei. One such method is the ab initio no-core shell model (NCSM). Starting from realistic two- and three-nucleon interactions this method can predict low-lying levels in p-shell nuclei. In this contribution, we present a brief overview of the NCSM with examples of recent applications. We highlight our study of the parity inversion in {sup 11}Be, for which calculations were performed in basis spaces up to 9{Dirac_h}{Omega} (dimensions reaching 7 x 10{sup 8}). We also present our latest results for the p-shell nuclei using the Tucson-Melbourne TM three-nucleon interaction with several proposed parameter sets.

Navratil, P; Ormand, W E; Forssen, C; Caurier, E

2004-11-30

128

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

SciTech Connect

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

Su, Bingjing; Hawari, Ayman, I.

2004-03-30

129

'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

130

Timescale Calculations for Ice Core Drilling Sites on the Temperate Ice Caps in Iceland  

NASA Astrophysics Data System (ADS)

Modelling of age vs. depth profiles and annual-layer thickness changes with depth in ice sheets forms part of the investigations carried out prior to the selection of ice core drilling sites. The well known Nye model, which assumes a constant vertical strain rate with depth in an ice sheet of thickness H is generally applicable in the upper half of polar and temperate ice caps, but the assumption of a constant vertical strain rate is unrealistic near the bed in an ice sheet frozen to bedrock. Dansgaard-Johnsen (D-J) type models assume that the vertical strain rate is constant down to height h above bedrock and then decreases linearly with depth towards zero at the bed. The parameter h can be calibrated according to the way in which the horizontal velocity varies with depth. Here we introduce a new derivation of the D-J model that accounts for bottom melting due to the geothermal heat flux, which averages 200 mW/m2 in Iceland. The model is then applied to five different locations on the temperate ice caps in Iceland, with ice thicknesses varying between 220 m and 850 m and accumulation rates ranging between 2.0 and 3.6 m ice/year. Data from ice cores drilled at three of these sites are used to calibrate the model. For the summit location on the Hofsjokull ice cap (H = 300 m), we find that a D-J model with a relatively high h/H ratio reproduces the timescale from a 100 m ice core better than the Nye model. Results indicate that a continuous precipitation record covering the last 400-500 years could be retrieved at the Hofsjokull summit (1790 m a.s.l.), and the assumption of bottom melting has a large effect on the modelled timescale at this site, yielding 50% lower ages at 90% of the ice depth than model runs that neglect bottom melting. For deeper drillings in Iceland, the ice-filled caldera at Bardarbunga, NW-Vatnajokull (H = 850 m), where a 415 m core was drilled in 1972, is among the most promising sites. Selection of the h/H ratio in the D-J model for timescale calculation within the caldera rims is complicated by an unusual ice-flow pattern but results strongly indicate that a 700-800 m ice core could yield a record covering historical time in Iceland (870 AD - present). Model results predict that by 90% of ice depth, the annual layers have thinned to 17 cm at the Hofsjokull summit and 8 cm within the Bardarbunga caldera. Annual layers of this thickness are detectable with the methods used in pilot ice core drilling and processing efforts in Iceland in recent years.

Thorsteinsson, T.; Einarsson, B.

2005-12-01

131

Cluster form factor calculation in the ab initio no-core shell model  

SciTech Connect

We derive expressions for cluster overlap integrals or channel cluster form factors for ab initio no-core shell model (NCSM) wave functions. These are used to obtain the spectroscopic factors and can serve as a starting point for the description of low-energy nuclear reactions. We consider the composite system and the target nucleus to be described in the Slater determinant (SD) harmonic oscillator (HO) basis while the projectile eigenstate to be expanded in the Jacobi coordinate HO basis. This is the most practical case. The spurious center of mass components present in the SD bases are removed exactly. The calculated cluster overlap integrals are translationally invariant. As an illustration, we present results of cluster form factor calculations for <{sup 5}He vertical bar{sup 4}He+n>, <{sup 5}He vertical bar{sup 3}H+d>, <{sup 6}Li vertical bar{sup 4}He+d>, <{sup 6}Be vertical bar{sup 3}He+{sup 3}He>, <{sup 7}Li vertical bar{sup 4}He+{sup 3}H>, <{sup 7}Li vertical bar{sup 6}Li+n>, <{sup 8}Be vertical bar{sup 6}Li+d>, <{sup 8}Be vertical bar{sup 7}Li+p>, <{sup 9}Li vertical bar{sup 8}Li+n>, and <{sup 13}C vertical bar{sup 12}C+n>, with all the nuclei described by multi-({Dirac_h}/2{pi}){omega} NCSM wave functions.

Navratil, Petr [Lawrence Livermore National Laboratory, L-414, P.O. Box 808, Livermore, California 94551 (United States)

2004-11-01

132

Criticality validation for burnup credit using recycle Pu criticals  

SciTech Connect

A set of 23 additional critical experiments were analyzed to provide additional input to the criticality validation portion of spent fuel cask analysis. The results of this analyses were combined with the previously analyzed criticals to determine the upper safety limit on k{sub eff}. The combined set of criticals can be used used for criticality validation for burnup credit, and are better suited for the range of isotopics in spent nuclear fuels. A trend observed in the analysis was that the calculated k{sub eff} deviates from the criticals in the positive direction, implying that increased burnup results in increased safety margin. 6 refs., 2 figs., 1 tab.

Fuentes, E.; Lancaster, D.

1997-04-01

133

Analysis of Mixed Oxide Fuel Loaded Cores in the Heavy Water Reactor FUGEN  

Microsoft Academic Search

Uranium-plutonium mixed oxide (MOX) fuel cores in the heavy water reactor, FUGEN, were analyzed using the Advanced Thermal Reactor (ATR) type core design code system WIMS-ATR\\/POLESTAR and the accuracy of this code system also has been evaluated by means of operational data through the 34 burnup cycles and on-site ?-scanning data. The root mean square errors of calculated thermal neutron

Tsukasa OHTANI; Takashi IIJIMA; Yoshitake SHIRATORI

2003-01-01

134

Analysis of Mixed Oxide Fuel Loaded Cores in the Heavy Water Reactor FUGEN  

Microsoft Academic Search

Uranium-plutonium mixed oxide (MOX) fuel cores in the heavy water reactor, FUGEN, were analyzed using the Advanced Thermal Reactor (ATR) type core design code system WIMS-ATR\\/POLESTAR and the accuracy of this code system also has been evaluated by means of operational data through the 34 burnup cycles and on-site ? -scanning data. The root mean square errors of calculated thermal

Tsukasa OHTANI; Takashi IIJIMA; Yoshitake SHIRATORI

2003-01-01

135

Characteristics of the WWR-K test core and the LEU LTAS to be placed in the central experimental beryllium device  

Microsoft Academic Search

In 2010 life test of three LEU (19.7%) lead test assemblies (LTA) is expected in the existing WWR-K reactor core with regular WWR-C-type fuel assemblies and a smaller core with a beryllium insert. Preliminary analysis of test safety is to be carried out. It implies reconstruction of the reactor core history for last three years, including burnup calculation for each

F. Arinkin; P. Chakrov; L. Chekushina; Gizatulin; S. Koltochnik; N. Hanan; P. Garner

2010-01-01

136

Irradiation Performance of Fast Reactor MOX Fuel Assemblies Irradiated to High Burnups  

Microsoft Academic Search

Mixed oxide fuel assemblies (MFA-1 and MFA-2 assemblies) were irradiated in the fast flux test facility to evaluate the irradiation performance of fast reactor core fuels at high burnups and high fast neutron fluences. The MFA-1 and MFA-2 assemblies achieved respective peak pellet burnups of 147 and 162GWd\\/t, and resisted to respective peak fast neutron fluences (E > 0:1 MeV)

Tomoyuki UWABA; Masahiro ITO; Tomoyasu MIZUNO

2008-01-01

137

Development of a Fully-Automated Monte Carlo Burnup Code Monteburns  

SciTech Connect

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

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

1999-01-01

138

From non-Hermitian effective operators to large-scale no-core shell model calculations for light nuclei  

Microsoft Academic Search

No-core shell model (NCSM) calculations using ab initio effective interactions are very successful in reproducing experimental nuclear spectra. The main theoretical approach is the use of effective operators, which include correlations left out by the truncation of the model space to a numerically tractable size. We review recent applications of the effective operator approach, within a NCSM framework, to the

Bruce R. Barrett; Ionel Stetcu; Petr Navrátil; James P. Vary

2006-01-01

139

On the use of effective core potentials in the calculation of magnetic properties, such as magnetizabilites and magnetic shieldings  

NASA Astrophysics Data System (ADS)

State-of-the art effective core potentials (ECPs) that replace electrons of inner atomic cores involve non-local potentials. If such an effective core potential is added to the Hamiltonian of a system in a magnetic field, the resulting Hamiltonian is not gauge invariant. This means, magnetic properties such as magnetisabilities and magnetic shieldings (or magnetic susceptibilities and nuclear magnetic resonance chemical shifts) calculated with different gauge origins are different even for exact solutions of the Schrödinger equation. It is possible to restore gauge invariance of the Hamiltonian by adding magnetic field dependent terms arising from the effective core potential. Numerical calculations on atomic and diatomic model systems (potassium mono-cation and potassium dimer) clearly demonstrate that the standard effective core potential Hamiltonian violates gauge invariance, and this affects the calculation of magnetisabilities more strongly than the calculation of magnetic shieldings. The modified magnetic field dependent effective core potential Hamiltonian is gauge invariant, and therefore it is the correct starting point for distributed gauge origin methods. The formalism for gauge including atomic orbitals (GIAO) and individual gauge for localized orbitals methods is worked out. ECP GIAO results for the potassium dimer are presented. The new method performs much better than a previous ECP GIAO implementation that did not account for the non-locality of the potential. For magnetic shieldings, deviations are clearly seen, but they amount to few ppm only. For magnetisabilities, our new ECP GIAO implementation is a major improvement, as demonstrated by the comparison of all-electron and ECP results.

van Wüllen, Christoph

2012-03-01

140

On the use of effective core potentials in the calculation of magnetic properties, such as magnetizabilites and magnetic shieldings.  

PubMed

State-of-the art effective core potentials (ECPs) that replace electrons of inner atomic cores involve non-local potentials. If such an effective core potential is added to the Hamiltonian of a system in a magnetic field, the resulting Hamiltonian is not gauge invariant. This means, magnetic properties such as magnetisabilities and magnetic shieldings (or magnetic susceptibilities and nuclear magnetic resonance chemical shifts) calculated with different gauge origins are different even for exact solutions of the Schro?dinger equation. It is possible to restore gauge invariance of the Hamiltonian by adding magnetic field dependent terms arising from the effective core potential. Numerical calculations on atomic and diatomic model systems (potassium mono-cation and potassium dimer) clearly demonstrate that the standard effective core potential Hamiltonian violates gauge invariance, and this affects the calculation of magnetisabilities more strongly than the calculation of magnetic shieldings. The modified magnetic field dependent effective core potential Hamiltonian is gauge invariant, and therefore it is the correct starting point for distributed gauge origin methods. The formalism for gauge including atomic orbitals (GIAO) and individual gauge for localized orbitals methods is worked out. ECP GIAO results for the potassium dimer are presented. The new method performs much better than a previous ECP GIAO implementation that did not account for the non-locality of the potential. For magnetic shieldings, deviations are clearly seen, but they amount to few ppm only. For magnetisabilities, our new ECP GIAO implementation is a major improvement, as demonstrated by the comparison of all-electron and ECP results. PMID:22443751

van Wüllen, Christoph

2012-03-21

141

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

142

The properties of iron under core conditions from first principles calculations  

Microsoft Academic Search

The Earth’s core is largely composed of iron (Fe). The phase relations and physical properties of both solid and liquid Fe are therefore of great geophysical importance. As a result, over the past 50 years the properties of Fe have been extensively studied experimentally. However, achieving the extreme pressures (up to 360GPa) and temperatures (?6000K) found in the core provide

L Vo?adlo; D. Alfè; M. J. Gillan; G. David Price

2003-01-01

143

Molecular dynamics calculation of liquid iron properties and adiabatic temperature gradient in the Earth's outer core  

Microsoft Academic Search

The knowledge of the temperature radial distribution in the Earth's core is important to understand the heat balance and conditions in the Earth's interior. Molecular dynamics (MD) simulations were applied to study the properties of liquid iron under the pressure-temperature conditions of the Earth's outer core. It is shown that the model used for the MD simulations can reproduce recent

L. Koci; A. B. Belonoshko; R. Ahuja

2007-01-01

144

Modeling of BH Loop for Core Loss Calculations in Power Transformer Using Finite Element Method  

Microsoft Academic Search

Users and manufacturers of transformers are nowadays capitalizing the core losses while considering the castings. Therefore, a software package which enables an engineer to predict quickly the approximate core loss in transformer of any rating and geometry will be very useful. The use of finite element methods for transformer design and analysis has been proven as a very powerful tool

K. Abbaszadeh; S. A. Gholamian; M. Ardebili; H. A. Toliyat

2006-01-01

145

Effect of a time varying power level in EBR-II on mixed-oxide fuel burnup  

Microsoft Academic Search

A refined prediction of burnup of mixed-oxide fuel in EBR-2 is compared with measured data. The calculation utilizes a time-varying power factor and results in a general improvement to previous calculations.

I. Z. Stone; J. W. Jost; R. B. Baker

1979-01-01

146

Investigation of Burnup Credit Issues in BWR Fuel  

SciTech Connect

Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel.

Broadhead, B.L.; DeHart, M.D.

1999-09-20

147

Hyperfine coupling of point defects in semiconductors by hybrid density functional calculations: The role of core spin polarization  

NASA Astrophysics Data System (ADS)

We implemented the calculation of hyperfine tensors into such plane wave supercell code working with the projector augmentation wave method that incorporates hybrid density functional theory and the contribution of the spin polarization of the core states. We show that the combination of HSE06 hybrid density functional together with the contribution of the core spin polarization provides accurate results on prominent point defects in various semiconductors, where the latter effect may be enormously large, in contrast to previous expectations. We briefly discuss the relevance of our results in the light of realization of solid-state quantum bits by paramagnetic point defects.

Szász, Krisztián; Hornos, Tamás; Marsman, Martijn; Gali, Adam

2013-08-01

148

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

149

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

150

Credit for burnup in spent-fuel storage rack design-regulatory perspective  

SciTech Connect

The motivation for taking credit for fuel assembly burnup in spent-fuel storage rack design is obvious. Several pressurized water reactor facilities, beginning with the standardized nuclear unit power plant system plants (Callaway and Wolf Creek) have been licensed to install racks that take credit for burnup. Designing racks to take credit for burnup introduces several calculational complexities with respect to criticality that are not present in designing for fresh fuel storage. The criticality analyses are required to be performed with calculational methods and procedures that have been verified by comparison with experiments. Several experiments have been performed with unirradiated fuel that included the major features of fuel racks. However, none has been performed with irradiated fuel. A final aspect of the two-region pool designs is assurance that only assemblies having the required burnup are stored in region 2.

Brooks, W.L.

1987-01-01

151

Dielectronic Recombination (via N=2-->N'=2 Core Excitations) and Radiative Recombination of Fe XX: Laboratory Measurements and Theoretical Calculations  

Microsoft Academic Search

We have measured the resonance strengths and energies for dielectronic recombination (DR) of Fe XX forming Fe XIX via N=2-->N'=2 (DeltaN=0) core excitations. We have also calculated the DR resonance strengths and energies using the AUTOSTRUCTURE, Hebrew University Lawrence Livermore Atomic Code (HULLAC), Multiconfiguration Dirac-Fock (MCDF), and R-matrix methods, four different state-of-the-art theoretical techniques. On average the theoretical resonance strengths

D. W. Savin; E. Behar; S. M. Kahn; G. Gwinner; A. A. Saghiri; M. Schmitt; M. Grieser; R. Repnow; D. Schwalm; A. Wolf; T. Bartsch; A. Müller; S. Schippers; N. R. Badnell; M. H. Chen; T. W. Gorczyca

2002-01-01

152

From Non-Hermitian Effective Operators to Large-Scale No-Core Shell Model Calculations for Light Nuclei  

SciTech Connect

No-core shell model (NCSM) calculations using ab initio effective interactions are very successful in reproducing experimental nuclear spectra. The main theoretical approach is the use of effective operators, which include correlations left out by the truncation of the model space to a numerically tractable size. We review recent applications of the effective operator approach, within a NCSM framework, to the renormalization of the nucleon-nucleon interaction, as well as scalar and tensor operators.

Barrett, B R; Stetcu, I; Navratil, P; Vary, J P

2006-03-06

153

BEAGL-01: a computer code for calculating rapid LWR core transients. Volume 1. Mathematical modeling. Computer code manual  

Microsoft Academic Search

BEAGL-01 is a computer program for calculating the conditions in a light water reactor core at steady state and while undergoing a transient. It couples a neutron kinetics model in R,Z geometry with a two-phase thermal-hydraulics model for multiple parallel channels. Appropriate applications include reactivity insertion accidents in both pressurized and boiling water reactors and transients in boiling water reactors

1984-01-01

154

Ab initio effective core potentials for molecular calculations. Potentials for the transition metal atoms Sc to Hg  

SciTech Connect

Ab initio effective core potentials (ECP's) have been generated to replace the Coulomb, exchange, and core-orthogonality effects of the chemically inert core electron in the transition metal atoms Sc to Hg. For the second and third transition series relative ECP's have been generated which also incorporate the mass--velocity and Darwin relativistic effects into the potential. The ab initio ECP's should facilitate valence electron calculations on molecules containing transition-metal atoms with accuracies approaching all-electron calculations at a fraction of the computational cost. Analytic fits to the potentials are presented for use in multicenter integral evaluation. Gaussian orbital valence basis sets are developed for the (3d,4s,4p), (4d,5s,5p), and (5d,6s,6p) orbitals of the first, second, and third transition series atoms, respectively. All-electron and valence-electron atomic excitation energies are also compared for the low-lying states of Sc--Hg, and the valence-electron calculations are found to reproduce the all-electron excitation energies (typically within a few tenths of an eV).

Hay, P.J.; Wadt, W.R.

1985-01-01

155

The burnup dependence of light water reactor spent fuel oxidation  

Microsoft Academic Search

The air oxidation of fragments of Light Water Reactor (LWR) spent fuel with burnup in the range 16-42 MWd\\/kg M was studied using thermogravimetric analysis. Experiments were conducted in dry air over the temperature range 255-325sp°C. Mass increases were generally followed until the calculated oxygen-to-metal ratio reached 2.7. LWR spent fuel was shown to oxidize via the two step reaction

Brady Dean Hanson

1998-01-01

156

Timescale Calculations for Ice Core Drilling Sites on the Temperate Ice Caps in Iceland  

Microsoft Academic Search

Modelling of age vs. depth profiles and annual-layer thickness changes with depth in ice sheets forms part of the investigations carried out prior to the selection of ice core drilling sites. The well known Nye model, which assumes a constant vertical strain rate with depth in an ice sheet of thickness H is generally applicable in the upper half of

T. Thorsteinsson; B. Einarsson

2005-01-01

157

Cluster form factor calculation in the ab initio no-core shell model  

Microsoft Academic Search

We derive expressions for cluster overlap integrals or channel cluster form factors for ab initio no-core shell model (NCSM) wave functions. These are used to obtain the spectroscopic factors and can serve as a starting point for the description of low-energy nuclear reactions. We consider the composite system and the target nucleus to be described in the Slater determinant (SD)

Petr Navrátil; Petr

2004-01-01

158

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

Microsoft Academic Search

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

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

159

Burnup credit issues in transportation and storage  

Microsoft Academic Search

Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the US experience in developing a methodology to implement burnup credit

M. C. Brady; T. L. Sanders; K. D. Seager; W. H. Lake

1992-01-01

160

Strategies for certifying a burnup credit cask  

Microsoft Academic Search

A new generation of high capacity spent fuel transport casks is being developed by the U.S. Department of Energy (DOE) as part of the Federal Waste Management System (FWMS). Burnup credit, which recognizes the reduced reactivity of spent fuel is being used for these casks. Both cask designs being developed for DOE burnup credit. The cask designs must be certified

W. H. Lake; J. R. Boshoven; L. A. Hassler

1992-01-01

161

Progress in Extending Burnup of LWR Fuel.  

National Technical Information Service (NTIS)

Progress in increasing burnup of LWR fuel has been and continues to be made. Initially, LWR fuels were designed to achieve a burnup of about 33 GWd/tu for PWRs and about 28 GWd/tu for BWRs. Current warranties are about 36 GWd/tu and 31 GWd/tu for PWRs and...

M. D. Freshley

1986-01-01

162

Ab initio no-core shell model calculations using realistic two- and three-body interactions  

Microsoft Academic Search

There has been significant progress in the ab initio approaches to the structure of light nuclei. One such method is the ab initio no-core shell model (NCSM). Starting from realistic two- and three-nucleon interactions this method can predict low-lying levels in p-shell nuclei. In this contribution, we present a brief overview of the NCSM with examples of recent applications. We

P. Navrátil; W. E. Ormand; C. Forssén; E. Caurier

2005-01-01

163

Calculation and verification of current limiting properties for Saturated Iron Core Superconductive Fault Current Limiter  

Microsoft Academic Search

This paper introduces a simulation model of Saturated Iron Core Superconducting Fault Current Limiter (SICSFCL) using SIMPOWER of MATLAB and establishes an actual model for the 35kV\\/90MVA SICSFCL installed in Puji substation at Kunming. The model is used to simulate the artificially imposed three-phase short-circuit and the results are compared to the live grid test results, which were carried out

Ziqiang Wei; Jingyin Zhang; Hui Hong; Weizhi Gong; Yin Xin

2011-01-01

164

Model biases in high-burnup fast reactor simulations  

SciTech Connect

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

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

2012-07-01

165

ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation  

SciTech Connect

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

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

2009-11-01

166

No-Core Shell Model Calculations in Light Nuclei with Three-Nucleon Forces  

SciTech Connect

The ab initio No-Core Shell Model (NCSM) has recently been expanded to include nucleon-nucleon (NN) and three-nucleon (3N) interactions at the three-body cluster level. Here it is used to predict binding energies and spectra of p-shell nuclei based on realistic NN and 3N interactions. It is shown that 3N force (3NF) properties can be studied in these nuclear systems. First results show that interactions based on chiral perturbation theory lead to a realistic description of {sup 6}Li.

Barrett, B R; Vary, J P; Nogga, A; Navratil, P; Ormand, W E

2004-01-08

167

Renormalization of power and burnup independent shape annealing function for excore detectors  

Microsoft Academic Search

This paper describes a predictive mathematical modeling of the excore neutron detectors using a renormalization of power and burnup independent shape annealing function for Korean Optimized Power Reactor (OPR-1000) and demonstrates its validity via comparison with the measurement data. The mathematical model is based on the assembly-wise spatial weighting functions and the core axial spatial weighting functions. Detector responses are

Moon-Ghu Park; Ho-Cheol Shin; Yu-Sun Choi; Dong-Hwan Park; Yonghee Kim

2009-01-01

168

FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP  

Microsoft Academic Search

OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be

Randy G. Lott

2003-01-01

169

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

170

Calculation and uncertainty analysis of 210 Pb dates for PIRLA project lake sediment cores  

Microsoft Academic Search

Lead-210 assay and dating are subject to several sources of error, including natural variation, the statistical nature of measuring radioactivity, and estimation of the supported fraction. These measurable errors are considered in calculating confidence intervals for 210Pb dates. Several sources of error, including the effect of blunders or misapplication of the mathematical model, are not included in the quantitative analysis.

Michael W. Binford

1990-01-01

171

Density functional theory calculation on many-cores hybrid central processing unit-graphic processing unit architectures  

NASA Astrophysics Data System (ADS)

We present the implementation of a full electronic structure calculation code on a hybrid parallel architecture with graphic processing units (GPUs). This implementation is performed on a free software code based on Daubechies wavelets. Such code shows very good performances, systematic convergence properties, and an excellent efficiency on parallel computers. Our GPU-based acceleration fully preserves all these properties. In particular, the code is able to run on many cores which may or may not have a GPU associated, and thus on parallel and massive parallel hybrid machines. With double precision calculations, we may achieve considerable speedup, between a factor of 20 for some operations and a factor of 6 for the whole density functional theory code.

Genovese, Luigi; Ospici, Matthieu; Deutsch, Thierry; Méhaut, Jean-François; Neelov, Alexey; Goedecker, Stefan

2009-07-01

172

Density functional theory calculation on many-cores hybrid central processing unit-graphic processing unit architectures.  

PubMed

We present the implementation of a full electronic structure calculation code on a hybrid parallel architecture with graphic processing units (GPUs). This implementation is performed on a free software code based on Daubechies wavelets. Such code shows very good performances, systematic convergence properties, and an excellent efficiency on parallel computers. Our GPU-based acceleration fully preserves all these properties. In particular, the code is able to run on many cores which may or may not have a GPU associated, and thus on parallel and massive parallel hybrid machines. With double precision calculations, we may achieve considerable speedup, between a factor of 20 for some operations and a factor of 6 for the whole density functional theory code. PMID:19624177

Genovese, Luigi; Ospici, Matthieu; Deutsch, Thierry; Méhaut, Jean-François; Neelov, Alexey; Goedecker, Stefan

2009-07-21

173

Calculation of the even harmonics of emf in the winding of a ring core magnetized by an external constant field, with magnetic hysteresis taken into account  

SciTech Connect

The calculation of the even harmonics of electromotive force in the secondary winding of a ferromagnetic ring core or a core extended in one direction, having a closed magnetic circuit in relation to the magnetic excitation flux and being magnetized in the plane of the core (ring) by a weak magnetic field, is carried out taking into account magnetic hysteresis and using the criterion of physical similarity.

Ponomarev, Yu.V.

1988-09-01

174

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

175

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

176

In-core thermal hydraulic and fission product calculations for severe fuel damage analyses  

SciTech Connect

Best-estimate calculations of realistic source terms are presented which reduce uncertainties in predicting fission product release from the UO{sub 2} fuel over the temperature range between 770 K and 3000 K. The proposed method of correlation includes such fuel morphology effects as equiaxed fuel grain growth and fuel-cladding interaction. The method correlates the product of fuel release rate and equiaxed grain size with the inverse fuel temperature to yield a bulk mass transfer correlation. It was found that less and slower releases are predicted utilizing the bulk mass transfer correlation than such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation. A Severe Fuel Damage (SFD) analysis code was developed to perform the thermal hydraulic and fission product calculations needed to analyze the Power Burst Facility SFD tests. The predictions utilizing the bulk mass transfer correlations overall followed the experimental time-release histories during the course of the heatup, power hold and cooldown phases of the transients. Good agreements were achieved for the integral releases both in timing and in magnitude. The proposed bulk mass transfer correlations can be applied to both current and advanced light water reactor fuels. 17 refs., 8 figs., 3 tabs.

Suh, K.Y.; Sharon, A.; Hammersley, R.J. (Fauske and Associates, Inc., Burr Ridge, IL (USA))

1989-05-01

177

PIUS Core Performance Analysis.  

National Technical Information Service (NTIS)

A detailed evaluation of the fuel burnup dependent power distribution and the scram reactivity for the PIUS reactor design has been performed. The analyses were carried out using the CPM lattice physics and NODE-P2 core neutronics/thermal-hydraulics codes...

J. F. Carew A. Aronson D. M. Cokinos A. Prince E. C. Selcow

1996-01-01

178

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

SciTech Connect

A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis.

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

1985-04-01

179

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

180

Calculation of core-level photoemission spectra of Mn films on Ag(001)  

NASA Astrophysics Data System (ADS)

We report a theoretical study of Mn 2p photoemission spectra for Mn thin films adsorbed on Ag(001), for which large satellites have recently been observed. Band structure calculations are performed to determine the ground-state electronic structure. The band structure information is then used to construct a realistic impurity model that includes full atomic multiplet interaction and coupling to the Mn-3d and Ag-4d bands. The model is applied to various Mn/Ag thin film structures, whereby only the most relevant model parameter, the hybridization strength, is varied using a simple scaling law. Good agreement with experiment is obtained for all systems. It is found that the satellite structure is due to the presence of (2p53d5) and (2p53d6L¯) final states, where the ligand hole L¯ is mainly in the majority spin Mn band of the neighbors. It is shown that atomic multiplet effects cause an apparent spin-orbit splitting that is greater for (2p53d5) than for (2p53d6) final states, which leads to a 1 eV larger main line to satellite splittings for the 2p1/2 lines than for the 2p3/2 lines in agreement with experiment.

Krüger, P.; Kotani, A.

2003-07-01

181

Core performance and proliferation resistance prospective of a novel natural uranium fueled, heavy water moderated nuclear research reactor  

Microsoft Academic Search

Three-dimensional burnup calculations were carried out to analyze the performance and proliferation resistance prospective of a novel natural uranium fueled, D2O moderated, D2O cooled and graphite reflected nuclear research reactor. The lattice simulation code WIMS-D\\/4 generated microscopic group constants, (?a, ?f, ?tr, ? etc.,), in conjunction with the diffusion theory based reactor core simulation code CITATION was employed in this

Mohammad Javed Khan; Aslam; Nasir Ahmad

2006-01-01

182

Recent advances in the practical and accurate calculation of core and valence XPS spectra of polymers: From interpretation to simulation?  

NASA Astrophysics Data System (ADS)

Core and valence X-ray Photoelectron Spectroscopies (XPS) are routinely used to obtain information on the chemical composition, bonding and homogeneity of polymer surfaces. In spite of their apparent conceptual simplicity, Core and Valence Electron Binding Energies (CEBEs and VEBEs) a few electron-volts (eV) or fractions of an eV apart are difficult to interpret. We present some results obtained with various recent theoretical approaches. An emphasis is made on a procedure based on the Density Functional Theory (DFT) that enables the calculation of CEBEs and VEBEs which are in remarkable agreement with experiment. The method has been tested on numerous small (3-6 atoms) to fairly large (15-25 atoms) molecules, and shows an average absolute deviation with experiment of only 0.20 eV for CEBEs and 0.30 eV for VEBEs, i.e. compatible with the resolution of the best XPS experiments carried out at the moment. Besides the quality of its predictions, the procedure takes advantage of the speed and CPU time scaling of DFT as a function of system size: it is computationally tractable, even for surprisingly large systems such as polymers, and may be an interesting accurate alternative to interpret and simulate XPS-probing on real systems. We illustrate the usefullness and pitfalls of this approach in fundamental as well as applied fields such as in the study of Polyacrylonitrile (PAN), Polytetrafluoroethylene (PTFE), Polyvinyldifluoride (PVdF) and ?-Aminopropyltriethoxysilane (?-APS, an adhesion promoter).

Bureau, C.; Lecayon, G.; Le Moël, A.; Chong, D. P.; Endo, K.; Delhalle, J.

1997-08-01

183

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

SciTech Connect

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

DeHart, M.D.

1996-05-01

184

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

185

Sensitivity and Uncertainty Analysis to Burn-up Estimates on ADS Using 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 reevaluation of some uncertainty XSs for ADS.

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

2005-02-11

186

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

187

The structures and properties of Si/SiO2 core/shell quantum dots studied by density-functional tight-binding calculations  

NASA Astrophysics Data System (ADS)

Si/SiO2 core/shell quantum dots (QDs) have been shown with wavelength-tunable photoluminescence in addition to their inert, nontoxic, abundant, low-cost, biocompatible advantages. Due to their big size, here, we apply density-functional tight-binding (DFTB) method to perform calculations to study their structures and properties. We systematically investigate the effects of surface passivation, thickness of SiO2 shell, and Si/O ratio on the structures and properties of Si/SiO2 core/shell quantum dots. We find that hydroxyl passivated Si/SiO2 core/shell quantum dots are able to stabilize the quantum dots compared with hydrogen passivated Si/SiO2 core/shell quantum dots. By using DFTB method, we are able to study Si/SiO2 core/shell quantum dots of big size (3 nm) and we find that, in Si/SiO2 core/shell quantum dots, there are competing effects between quantum confinement (blueshift) and oxidation (redshift) with the decrease of the size of Si core. The transition point is when Si/SiO2 ratio is around 1:1. The effect of the thickness of SiO2 on energy gap is not as significant as the effect of the size of the Si core. Our study provides theoretical basis for designing Si quantum dots with tunable photoluminescence.

Dong, Huilong; Hou, Tingjun; Sun, Xiaotian; Li, Youyong; Lee, Shuit-Tong

2013-09-01

188

Use of burnup credit for transportation and storage  

Microsoft Academic Search

Burnup credit is the application of the effects of fuel burnup to nuclear criticality design. When burnup credit is considered in the design of storage facilities and transportation casks for spent fuel, the objectives are to reduce the requirements for storage space and to increase the payload of casks with acceptable nuclear criticality safety margins. The spent-fuel carrying capacities of

T. L. Sanders; R. I. Ewing; W. H. Lake

1991-01-01

189

Chemical Reprocessing of Fuels with High Burnup in France.  

National Technical Information Service (NTIS)

Experience of the CEA in reprocessing fuels with high burnup has been acquired on: oxides of enriched uranium with burnup of 20,000 MWd/t in the EL3 reactor (aqueous process--Fontenay), UO sub 2 --PuO sub 2 mixed oxide with burnup of 45,000 MWd/t (aqueous...

J. Sauteron P. Faugeras

1969-01-01

190

On the importance of thermodynamic self-consistency for calculating clusterlike pair correlations in hard-core double Yukawa fluids  

NASA Astrophysics Data System (ADS)

Understanding the mechanisms of clustering in colloids, nanoparticles, and proteins is of significant interest in material science and both chemical and pharmaceutical industries. Recently, using an integral equation theory formalism, Bomont et al. [J. Chem. Phys. 132, 184508 (2010)] studied theoretically the temperature dependence, at a fixed density, of the cluster formation in systems where particles interact with a hard-core double Yukawa potential composed of a short-range attraction and a long-range repulsion. In this paper, we provide evidence that the low-q peak in the static structure factor, frequently associated with the formation of clusters, is a common behavior in systems with competing interactions. In particular, we demonstrate that, based on a thermodynamic self-consistency criterion, accurate structural functions are obtained for different choices of closure relations. Moreover, we explore the dependence of the low-q peak on the particle number density, temperature, and potential parameters. Our findings indicate that enforcing thermodynamic self-consistency is the key factor to calculate both thermodynamic properties and static structure factors, including the low-q behavior, for colloidal dispersions with both attractive and repulsive interactions. Additionally, a simple analysis of the mean number of neighboring particles provides a qualitative description of some of the cluster features.

Kim, Jung Min; Castañeda-Priego, Ramón; Liu, Yun; Wagner, Norman J.

2011-02-01

191

On the importance of thermodynamic self-consistency for calculating clusterlike pair correlations in hard-core double Yukawa fluids.  

PubMed

Understanding the mechanisms of clustering in colloids, nanoparticles, and proteins is of significant interest in material science and both chemical and pharmaceutical industries. Recently, using an integral equation theory formalism, Bomont et al. [J. Chem. Phys. 132, 184508 (2010)] studied theoretically the temperature dependence, at a fixed density, of the cluster formation in systems where particles interact with a hard-core double Yukawa potential composed of a short-range attraction and a long-range repulsion. In this paper, we provide evidence that the low-q peak in the static structure factor, frequently associated with the formation of clusters, is a common behavior in systems with competing interactions. In particular, we demonstrate that, based on a thermodynamic self-consistency criterion, accurate structural functions are obtained for different choices of closure relations. Moreover, we explore the dependence of the low-q peak on the particle number density, temperature, and potential parameters. Our findings indicate that enforcing thermodynamic self-consistency is the key factor to calculate both thermodynamic properties and static structure factors, including the low-q behavior, for colloidal dispersions with both attractive and repulsive interactions. Additionally, a simple analysis of the mean number of neighboring particles provides a qualitative description of some of the cluster features. PMID:21322731

Kim, Jung Min; Castañeda-Priego, Ramón; Liu, Yun; Wagner, Norman J

2011-02-14

192

OCTOPUS burnup and criticality code system.  

National Technical Information Service (NTIS)

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate bu...

J. L. Kloosterman J. C. Kuijper P. F. A. Leege

1996-01-01

193

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

194

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 cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs.

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

1986-01-01

195

Assessment of on-line burnup monitoring of pebble bed reactor fuel using passive gamma-ray spectrometry  

Microsoft Academic Search

An investigation was performed to assess the feasibility of passive gamma-ray spectrometry assay as an approach for on-line burnup determination for the Modular Pebble Bed Reactor (MPBR). In addition to its inherently safe design, a unique feature of this reactor is its multipass fuel cycle in which graphite fuel pebbles are randomly loaded and continuously circulated through the core until

Ayman I. Hawari; Jianwei Chen; Bingjing Su; Zhongxiang Zhao

2002-01-01

196

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

SciTech Connect

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

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

2006-07-01

197

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

198

Variational approach to the calculation of the radii in the stagnant core model of shaped charge jet formation  

Microsoft Academic Search

The formation of a jet and slug from a collapsing shaped charge liner can be viewed, by appropriate choice of reference frame, as the result of two fluid streams impinging upon each other. In this article we consider this formation process and develop further the concept of a stagnant core model. In this model a core region of material is

J. P. Curtis; R. J. Kelly; K. G. Cowan

1994-01-01

199

A Tight Lattice, Epithermal Core Design for the Integral PWR  

SciTech Connect

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

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

2004-07-01

200

The Encapsulated Nuclear Heat Source (ENHS) Reactor Core Design  

Microsoft Academic Search

A once-for-life, uniform composition, blanket-free and fuel-shuffling-free reference core has been designed for the Encapsulated Nuclear Heat Source (ENHS) to provide the design goals of a nearly zero burnup reactivity swing throughout â20 yr of full-power operation up to the peak discharge burnup of more than 100 GWd\\/t HM. What limits the core life is radiation damage to the HT-9

Ser Gi Hong; Ehud Greenspan; Yeong Il Kim

2005-01-01

201

Recent advances in the practical and accurate calculation of core and valence XPS spectra of polymers: from interpretation to simulation?  

Microsoft Academic Search

Core and valence X-ray Photoelectron Spectroscopies (XPS) are routinely used to obtain information on the chemical composition, bonding and homogeneity of polymer surfaces. In spite of their apparent conceptual simplicity, Core and Valence Electron Binding Energies (CEBEs and VEBEs) a few electron-volts (eV) or fractions of an eV apart are difficult to interpret. We present some results obtained with various

Christophe Bureau; Delano P. Chong; Kazunaka Endo; Joseph Delhalle; Gérard Lecayon; Alain Le Moël

1997-01-01

202

Atmospheric burnup of the cosmos-954 reactor.  

PubMed

On 24 January 1978 the Russian satellite Cosmos-954 reentered the atmosphere over northern Canada. By use of high-altitude balloons, the atmosphere was sampled during 1978 up to an altitude of 39 kilometers to detect particulate debris from the reactor on board the satellite. Enriched uranium-bearing aerosols at concentrations and particle sizes compatible with partial burnup of the Cosmos-954 reactor were detected only in the high polar stratosphere. PMID:17729681

Krey, P W; Leifer, R; Benson, W K; Dietz, L A; Hendrikson, H C; Coluzza, J L

1979-08-10

203

Value of burnup credit beyond actinides  

Microsoft Academic Search

DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping

D. Lancaster; E. Fuentes; Chi Kang

1997-01-01

204

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

205

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

NASA Astrophysics Data System (ADS)

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

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

2005-05-01

206

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

SciTech Connect

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

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

2005-05-24

207

Two-component relativistic density-functional calculations of the dimers of the halogens from bromine through element 117 using effective core potential and all-electron methods  

Microsoft Academic Search

A two-component quasirelativistic Hamiltonian based on spin-dependent effective core potentials is used to calculate ionization energies and electron affinities of the heavy halogen atom bromine through the superheavy element 117 (eka-astatine) as well as spectroscopic constants of the homonuclear dimers of these atoms. We describe a two-component Hartree-Fock and density-functional program that treats spin-orbit coupling self-consistently within the orbital optimization

Alexander V. Mitin; Christoph van Wüllen

2006-01-01

208

Effects of high burnup on spent-fuel casks  

Microsoft Academic Search

Utility fuel managers have become very interested in higher burnup fuels as a means to reduce the impact of refueling outages. High-burnup fuels have significant effects on spent-fuel storage or transportation casks because additional heat rejection and shielding capabilities are required. Some existing transportation casks have useful margins that allow shipment of high-burnup fuel, especially the NLI-1\\/2 truck cask, which

1986-01-01

209

Impact of High Burnup on PWR Spent Fuel Characteristics  

SciTech Connect

Reducing the burden of management of spent nuclear fuel is important to the future of nuclear energy. The impact of higher pressurized water reactor (PWR) fuel burnup is examined in this paper from the perspective of its impact on spent-fuel radioactivity, decay heat, and plutonium content. The necessary fresh fuel enrichments to achieve high burnup in PWRs with the same three-batch operation scheme are first computed; then, characteristics of the spent fuel are determined. The increase in decay heat with burnup is found to be generally less than linear. Although each high-burnup fuel assembly would be hotter and more radioactive, the total decay heat to be removed or accommodated in storage is less for the same electricity production. If the time window before 150 yr after discharge can be excluded from impacting a repository, significant savings in its capacity can be realized with high-burnup fuel. The high-burnup fuel is more proliferation resistant because of reduced total plutonium production per kilowatt hour and because of higher content of less desirable plutonium isotopes, such as {sup 238}Pu. The fuel cycle cost can be slightly reduced by increasing burnup until it reaches a shallow minimum near 70 MWd/kg. Higher burnups would require one-time changes to the limits on enrichments that can be handled in most commercial fuel fabrication facilities. Changing the waste fee to base it on the amount of radioactivity in the spent fuel would enhance the economic benefit of high burnup.

Xu Zhiwen; Kazimi, Mujid S.; Driscoll, Michael J. [Massachusetts Institute of Technology (United States)

2005-11-15

210

Results from an international intercomparison of fundamental mode benchmark calculations for steam ingress into gas-cooled fast reactor cores  

Microsoft Academic Search

Steam ingress into a gas-cooled fast reactor (GCFR) core may lead to reactivity effects that are undesirable from the point of view of reactor safety. Unfortunately, the amount of reactivity increase caused by a certain steam concentration is usually subject to considerable uncertainty, as has become evident by occasional comparisons between various laboratories for specific examples. Therefore, some time ago,

Kiefhaber

1982-01-01

211

Coupling of mantle temperature anomalies and the flow pattern in the core: interpretation based on simple convection calculations  

NASA Astrophysics Data System (ADS)

The association of some relatively stationary features of the magnetic field at the core-mantle boundary (CMB) with temperature anomalies in the deep mantle inferred from seismic tomography has led to the speculation that flow in the core is locked to thermal anomalies in the mantle, with downwellings (upwellings) in the core lying below cold (hot) anomalies in the lower mantle (Bloxham and Gubbins). The relatively large temperature anomalies in the mantle act to control the heat flux out of the core, with high heat flux into colder than average mantle and low (perhaps negative) heat flux into regions of hotter than average mantle. We use a simple two-dimensional, internally heated convection model in a Cartesian geometry with varying thermal boundary conditions to investigate this problem. The relationship between upwellings (downwellings) and maxima (minima) in surface heat flux depends upon the thermal boundary condition, with maximum heat flux sometimes associated with upwellings, sometimes downwellings, and sometimes neither. However, for moderate Rayleigh numbers, large aspect ratio geometries and large variations in heat flux, the Bloxham and Gubbins hypothesis is supported.

King, Scott D.; Hager, Bradford H.

1989-12-01

212

Variational approach to the calculation of the radii in the stagnant core model of shaped charge jet formation  

NASA Astrophysics Data System (ADS)

The formation of a jet and slug from a collapsing shaped charge liner can be viewed, by appropriate choice of reference frame, as the result of two fluid streams impinging upon each other. In this article we consider this formation process and develop further the concept of a stagnant core model. In this model a core region of material is supposed to be stationary at the junction where the liner material turns to form the jet and slug. In our two-dimensional treatment the boundaries of the core region and the free streamlines are assumed to be arcs of circles and the main problem is to determine the radii of these boundaries. However, unlike in previous work, a nonuniform flow field is assumed to exist in the circular flow region from the outset. The nonuniform flow field we derive needs to be matched with the (assumed) uniform flow in the impinging stream. To accomplish this a transition region in the impinging stream is postulated. Consideration of the mass and momentum balances in this region leads to further model equations. The first of these balances gives a relation between the radii of the free streamline and the stagnant core boundary. It is shown that there are no physically acceptable exact solutions to the model equations when the energy is minimized. However a very accurate approximate solution is shown to exist. This solution leads to an expression for the liner speed on the core boundary which is identical to the critical speed used in a recent study on the formation of incoherent jets. Physically sensible values of the free streamline radius are also shown to result from this approximate solution.

Curtis, J. P.; Kelly, R. J.; Cowan, K. G.

1994-12-01

213

I. Core-quasiparticle coupling model calculations as a test of IBA core descriptions of the even-mass Hg isotopes. II. Decay of mass-separated /sup 203/At  

SciTech Connect

The Core-Quasiparticle Coupling Model (CQCM) for odd-mass nuclei, which is based on dynamical field theory and the Bardeen-Cooper-Schrieffer (BCS) method, has been applied to two problems. In the first, a study of Pauli exchange effects for the odd particle in the Interacting Boson-Fermion Model (IBFM) is made by a comparison with the CQCM. Spectra for a partly filled j-shell coupled to Interacting Boson Model cores are calculated in both models, and values of the IBFM exchange parameter ..lambda../sub 0/ are found which lead to spectra similar to those in the CQCM. In the second application, CQCM calculations have been performed for the i/sub 13/2/ bands in the odd-mass Hg isotopes, h/sub 9/2/ bands in the odd-mass Tl isotopes, and the h/sub 11/2/ bands in the odd-mass Au isotopes (A = 189 to 195). The even-mass Hg core descriptions were taken from previous proton-neutron Interacting Boson Model (IBM-2) calculations, and thus a comparison between the experimental and calculated level schemes for the odd-mass isotopes can be interpreted as a test of the core description. A predicted transition from positive Q(2/sup +/) near A = 195 to negative Q(2/sup +/) near A = 191 is not borne out by the odd-mass data. The radioactive decay of /sup 203/At to /sup 203/Po has been studied with mass-separated sources from the UNISOR facility. Time-sequenced spectra of ..gamma..-rays, x-rays, and conversion electrons have been obtained, together with ..gamma..-..gamma..-t, ..gamma..-X-t, e/sup -/-..gamma..-t and e/sup -/-X-t coincidence data. From this information, a decay scheme has been constructed, consisting of 30 excited states and 45 transitions that incorporate approximately 90% of the decay intensity assigned to /sup 203/At. All excited states below 1 MeV have been assigned unique spin-parity values, and the observed level scheme can be qualitatively understood in terms of a particle-core weak coupling description. The ground-state decay energy has been deduced.

Semmes, P.B.

1985-07-01

214

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

SciTech Connect

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

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

2010-01-01

215

Analyzing the rod drop accident in a BWR with high burnup fuel  

Microsoft Academic Search

The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. Calculations of this design-basis event has been done conservatively because there was margin to the fuel failure criterion of 170 cal\\/g. Because high burnup fuel may fail at much lower fuel enthalpies, the best-estimate of the enthalpy and

D. J. Diamond; L. Neymotin

1997-01-01

216

Value of burnup credit beyond actinides  

Microsoft Academic Search

The U.S. Department of Energy has submitted a topical report to the U.S. Nuclear Regulatory Commission (NRC) justifying burnup credit based only on actinide isotopes (²³⁴U, ²³⁵U, ²³⁶U, ²³⁸U, ²³⁸Pu, ²³⁹Pu, ²⁴°Pu, ²⁴¹Pu, ²⁴²Pu, and ²⁴¹Am). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel (SNF) to be transported in significantly

D. Lancaster; E. Fuentes; C. Kang

1997-01-01

217

Heterometallic clusters with a new {Re3Mo3S8} core: direct synthesis, properties and DFT calculations.  

PubMed

We report the direct facile high-temperature synthesis and investigation of heterometallic rhenium-molybdenum cluster K6[Re3Mo3S8(CN)4(CN)2/2] () with a new {Re3Mo3S8} core. Dissolution of polymeric compound resulted in subsequent oxidation and formation of stable 23e paramagnetic anionic cluster complex [Re3Mo3S8(CN)6](6-). PMID:24045787

Gayfulin, Yakov M; Naumov, Nikolay G; Rizhikov, Maxim R; Smolentsev, Anton I; Nadolinny, Vladimir A; Mironov, Yuri V

2013-10-01

218

First-principles calculational methods for surface-vacancy formation energies, heats of segregation, and surface core-level shifts  

Microsoft Academic Search

The matrix Green's-function version of the self-consistent scattering theory of surface-point-defect energetics is extended to the cases of vacancy formation, substitutional adsorption, and surface core-level shifts. Vacancies are treated by zeroing Hamiltonian and overlap matrix elements involving orbitals of atoms being removed and crystal orbitals or the crystal potential. Substitutional adsorption is treated as ordinary adsorption in a vacancy, a

Peter Feibelman

1989-01-01

219

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

220

Verification of the ORIGEN2 code analysis for the TMI-2 reactor core  

SciTech Connect

Accurate definition of the fission product inventories produced in the TMI-2 reactor prior to the accident on March 29, 1979 are of considerable interest to many organizations including the Department of Energy which is shipping the damage reactor core to the Idaho National Engineering Laboratory and conducting the TMI-2 reactor examination program, and General Public Utilities which is defueling the reactor. Numerous fission product inventory calculations have been performed for the TMI-2 core, including an ORIGEN2 analysis by EG G which uses 1239 nodes to define burnup in the reactor core. To provide a verification of the predicted fission product inventories, a measurement study was performed using pellets from various core regions. Measurements were performed for transuranics, burnup monitors, noble gases, principal gamma ray emitters, {sup 129}I, and {sup 90}Sr. Comparisons between the experimental results and the code analyses are presented with an evaluation of the associated uncertainties. Also, a discussion is presented of the probable causes of the observed differences between the code and measured values. 10 refs., 1 fig., 2 tabs.

Akers, D.W.; Schnitzler, B.G.

1988-01-01

221

Core materials development for the fuel cycle R&D program  

Microsoft Academic Search

The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels’ fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300dpa

S. A. Maloy; M. Toloczko; James I. Cole; Thak Sang Byun

2011-01-01

222

Core materials development for the fuel cycle R&D program  

Microsoft Academic Search

The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels' fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300

S. A. Maloy; M. Toloczko; J. Cole; T. S. Byun

2011-01-01

223

Overview of effects of burnup credit on cask design  

Microsoft Academic Search

A number of opportunities exist to increase the productivity of the next generation of spent fuel shipping casks. One of the opportunities being evaluated by Sandia National Laboratories (SNL) under the sponsorship of the Department of Energy's (DOE's) Cask System Development Program at Idaho is the implementation of burnup credit in the design of spent fuel casks. Burnup credit is

1988-01-01

224

Vortices of polarization in BaTiO3 core-shell nanoceramics: Calculations based on ab initio derived Hamiltonian versus Landau theory  

NASA Astrophysics Data System (ADS)

In this paper, we want to emphasize the fact that many experimental properties of ceramics can be explained by the existence of a core-shell structure of the grains, particularly at small sizes. In this framework, we have studied BaTiO3 (BT) ceramics constituted of core-shell nanoparticles, nanowires, or nanoplanes by using ab initio derived effective Hamiltonian calculations whose application range is for large values of shell thickness and low values of shell permittivity. Many differences and new features compared to the situation of nanodots are induced by the core-shell structure. For instance, phase sequences are different; there is also a coexistence of vortices found by Naumov, Bellaiche, and Fu [I. I. Naumov, L. Bellaiche, and H. Fu, Nature (London)10.1038/nature03107 432, 737 (2004)] in the case of isolated dots with a homogeneous polarization, a transition from cubic paraelectric phase towards nonpolar rhombohedral phase, anomalies in dielectric permittivity associated with the onset of toroidal moments, etc. Afterwards, we compare these results with those obtained by the Landau theory of core-shell ceramics we have recently published. However, the ab initio calculations fail to capture the physics at small shell thickness and/or high shell permittivity, whereas the Landau theory fails to predict the peculiar properties of the phases in which vortices exist. Therefore, in a tentative way to build a global theory, we have constructed a Landau potential using both the polarization and the toroidal moment as competing order parameters, which allows us to propose a phase diagram, whatever the thickness and permittivity of the shell are.

Anoufa, M.; Kiat, J. M.; Kornev, I.; Bogicevic, C.

2013-10-01

225

Isotopic validation for PWR actinide-only burnup credit using Mihama-3 data  

SciTech Connect

This report augments a US Department of Energy Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages submitted to the Nuclear Regulatory Commission (NRC). The purpose of the topical report was to obtain NRC approval on a generic burnup credit methodology. A major part of the methodology is the validation of the neutronics model used for designing the criticality control system. The validation methodology presented in the topical report established isotopic correction factors based on 19 isotopic samples. This report presents additional data points for isotopic validation. Measured and calculated isotopic quantities for nine samples from three spent fuel assemblies discharged from the Japanese Mihama-3 reactor are tabulated. 5 refs., 2 tabs.

Rahimi, M.; Lancaster, D.

1996-10-01

226

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

Microsoft Academic Search

The ACO-3 irradiation test, which attained extremely high burnups of about 232GWd\\/t and resisted a high neutron fluence (E>0.1MeV) of about 39×1026n\\/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic\\/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of

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

2011-01-01

227

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

228

The zonal tidal effect on the variation in the rotation rate of the Earth with a fluid core II. Numerical calculation and comparisons  

NASA Astrophysics Data System (ADS)

The tidal variation in Earth rotation rate is a periodical response to solar-lunar tide generating potential (TGP). Some theoretical formulae are given here based on Doodson development of TGP including the variations in Earth rotation rate, LOD and UT1. Finally the zonal tidal effect on the variation in the fluid core Earth rotation rate is calculated according to the formula deduced by Xi Qinwen (1995). The calculation shows that the results in this paper are well consistent with the ones in IERS (96), which indicates the correctness of the theoretical formula we deduced. It is also shown that the effects from the high frequency parts are relatively small, within the observing precision so far; relatively large effects due to the lower parts, which should be able to be seperated from the observed data, are actually difficult to make because of the influence from some non-tidal factors as well as short time span data.

Zhang, Han-Wei; Zheng, Yong; Du, Lan; Pan, Guan-Song

229

Correlated, relativistic, and basis set limit molecular polarizability calculations to evaluate an augmented effective core potential basis set  

NASA Astrophysics Data System (ADS)

Initial investigations have demonstrated that an augmented ECP basis set can be used to calculate valence electronic properties with deviations of less than 1% from all-electron basis sets. Past work has largely focused on molecules with relatively light atoms (Z<18) examined with time-dependent Hartree-Fock (TDHF) theory. In this work, the dipole moment and polarizability of a number of well-studied molecules are examined with HF, MP2, CCSD, and CCSD(T) correlated wave functions. Additionally, systems not as thoroughly studied due to the difficulty of all-electron calculations when Z=50-85 are included. The SBK ECP basis set, augmented with optimized valence functions, performs well across a broad range of methods, less than 3% different from all electron relativistic and correlated wave functions. Orders of magnitude time savings (101-104) are exchanged for a minimal difference from all-electron basis sets.

Labello, Nicholas P.; Ferreira, Antonio M.; Kurtz, Henry A.

230

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

231

Analyzing the rod drop accident in a BWR with high burnup fuel  

SciTech Connect

The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. Calculations of this design-basis event has been done conservatively because there was margin to the fuel failure criterion of 170 cal/g. Because high burnup fuel may fail at much lower fuel enthalpies, the best-estimate of the enthalpy and the uncertainty is of interest. In part of this study, calculations assessed the sensitivity to reactor conditions such as control rod pattern, inlet subcooling, and fuel burnup. It was shown that fuel enthalpy at any location in the region surrounding the dropped rod depends on the rod worth, the distance from the dropped rod, and the burnup of the fuel. The study also calculated the sensitivity to parameters whose modeling introduces significant uncertainty which may increase with burnup. These parameters are the control rod worth, Doppler reactivity coefficient, delayed neutron precursor fraction, and fuel specific heat. The results of the sensitivity studies were used in a model to determine the random uncertainty in the fuel enthalpy. The standard deviation for the calculated fuel enthalpy was estimated to be 37%. Therefore, the limiting bundle fuel enthalpy might be 75% higher than calculated. The effect of the fuel rod enthalpy distribution within a bundle was also investigated. RAMONA-4B calculates the fuel bundle average enthalpy and estimates must be made of a bundle peaking factor to determine the fuel rod enthalpy. A fit of RAMONA-4B bundle powers was used to estimate the local power peaking. It was determined that the peaking factor could be 25% higher than the factor usually assumed for RDA analysis. Combining this error with the random error means that for this analysis the actual fuel rod enthalpy could be 100% larger than calculated by RAMONA-4B. This is much larger than the uncertainty in most parameters calculated with best-estimate methods for design-basis events. 10 refs., 3 figs.

Diamond, D.J.; Neymotin, L.

1997-02-01

232

LETTER TO THE EDITOR: R-matrix calculations of core-excited negative-ion states in Li  

Microsoft Academic Search

The excitation cross sections for the lowest-lying autoionizing states 1s2s2 2S, 1s2s2p 4P and 1s(2s2p 3P) 2P in lithium have been calculated by the R-matrix method with pseudo-states. The close-coupling expansion includes 37 target states with the 1s22l, 1s23l, 1s2l2l´, 1s2l3l´ configurations in the LS-coupling scheme. The excitation cross sections reveal a few prominent resonances in the near-threshold regions. The

Oleg Zatsarinny

1999-01-01

233

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

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

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

2011-07-01

234

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

SciTech Connect

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

Conde, J.M.; Recio, M. [Consejo de Seguridad Nuclear, Madrid (Spain)

1995-04-01

235

High-accuracy MR-MP Perturbation Theory Energy and Radiative Rates Calculations for Core-excited Transitions in Fe XVI  

NASA Astrophysics Data System (ADS)

Accurate theoretical energy level, lifetime, and transition probability calculations of core-excited Fe XVI were performed employing the relativistic Multireference Møller-Plesset perturbation theory. In these computations the term energies of the highly excited n <= 5 states arising from the configuration 1s 22sk 2pm 3l p nl' 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.

Díaz, F.; Vilkas, M. J.; Ishikawa, Y.; Beiersdorfer, P.

2013-07-01

236

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

237

Pancake core high conversion light water reactor concept  

Microsoft Academic Search

A new concept is proposed for a high conversion light water reactor (HCLWR) that achieves both high conversion and high burnup while maintaining a negative void reactivity coefficient. This HCLWR has a flat pancake core with thick axial blankets. By using the flat core, a potential problem of HCLWRs, the positive void reactivity coefficient can be reduced by neutron leakage,

Y. Ishiguro; K. Okumura

1989-01-01

238

Band-edge ultrafast pump-probe spectroscopy of core/shell CdSe/CdS rods: assessing electron delocalization by effective mass calculations.  

PubMed

CdSe/CdS dot/rods nanocrystals show interesting physical properties related to the band-alignment at the hetero-interface, which controls the band-edge electron delocalization over the rods. Here the differential transmission spectra of CdSe/CdS nanorod samples with different core sizes have been measured using excitation resonant to the core transition. The photo bleaching ratio between dot and rod transitions increases with the dot size, indicating a trend towards electron localization. This trend has been further quantified by performing effective mass calculations in which the conduction band misalignment was varied in order to reproduce the observed bleaching feature ratio. The best agreement was found for negligible conduction band misalignment for small dots of around 2.3 nm in diameter, and about -0.1 eV misalignment was estimated for the larger dots, above 3.5 nm in diameter. This shows that the band misalignment might be dependent on the geometry of the system, and we argue that this might be related to different strain developed at the hetero-interface. PMID:22523752

Lupo, Maria Grazia; Scotognella, Francesco; Zavelani-Rossi, Margherita; Lanzani, Guglielmo; Manna, Liberato; Tassone, Francesco

2012-04-20

239

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

SciTech Connect

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

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

1987-10-01

240

Comparison of experimentally determined spent-fuel compositions with ORIGEN 2 calculations  

SciTech Connect

The specific experimental measurements of interest here involve the determination of parameters related to the actinide and fission product composition of samples from five elements taken from fuel assemblies discharged from the Turkey Point Unit 3 PWR. Two fuel assemblies were obtained for the purposes of nondestructive and destructive assay. These assemblies were initially fueled with 448 kg of UO/sub 2/ enriched to 2.556 wt % /sup 235/U and then irradiated for 851 full-power days. Five elements were then removed, and assay samples were taken from each element near the core midplane. The relevant parameters measured were /sup 148/Nd//sup 238/U, /sup 239/Pu//sup 238/U, and the isotopic compositions of U, Pu, Kr, and Xe. Fuel depletion calculations were performed using the updated ORIGEN2 PWR model. The burnup of the fuel was determined by adjusting the ORIGEN2 fuel burnup to match the experimentally determined /sup 148/Nd//sup 238/U ratio for each fuel element. The resulting burnup was then used to calculate the other experimentally determined parameters listed above. The agreement between ORIGEN2 and the experimental results is very good, with the average error for five samples being < 4% for most parameters. Based on this comparison, it appears that the ORIGEN2 computer code is capable of accurately calculating the composition of irradiated fuel from a modern PWR. However, well-characterized experimental measurements should continue to be obtained for validation purposes because the calculated values of many nuclides, particularly the minor actinides, still have significant uncertainties.

Croff, A.G.

1981-01-01

241

Two-component relativistic density-functional calculations of the dimers of the halogens from bromine through element 117 using effective core potential and all-electron methods.  

PubMed

A two-component quasirelativistic Hamiltonian based on spin-dependent effective core potentials is used to calculate ionization energies and electron affinities of the heavy halogen atom bromine through the superheavy element 117 (eka-astatine) as well as spectroscopic constants of the homonuclear dimers of these atoms. We describe a two-component Hartree-Fock and density-functional program that treats spin-orbit coupling self-consistently within the orbital optimization procedure. A comparison with results from high-order Douglas-Kroll calculations--for the superheavy systems also with zeroth-order regular approximation and four-component Dirac results--demonstrates the validity of the pseudopotential approximation. The density-functional (but not the Hartree-Fock) results show very satisfactory agreement with theoretical coupled cluster as well as experimental data where available, such that the theoretical results can serve as an estimate for the hitherto unknown properties of astatine, element 117, and their dimers. PMID:16483205

Mitin, Alexander V; van Wüllen, Christoph

2006-02-14

242

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

243

Whole-core LEU fuel demonstration in the ORR  

SciTech Connect

A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U/sub 3/Si/sub 2/ at 4.8 Mg U/m/sup 3/ and shim rod fuel followers will contain U/sub 3/Si/sub 2/ at 3.5 Mg U/m/sup 3/. Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U/sub 3/Si/sub 2/ fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worths, cycle length, fuel discharge burnup, gamma heating rates, ..beta../sub eff/l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed.

Snelgrove, J.L.; Bretscher, M.M.; Cornella, R.J.; Hobbs, R.W.

1985-01-01

244

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

245

Neutron and Gamma Ray Source Evaluation of LWR High Burn-up UO2and MOX Spent Fuels  

Microsoft Academic Search

The axial neutron emission and gamma ray source distribution were measured for LWR high burn-up UO2 and MOX spent fuel rods. The gamma rays of Cs, Cs and Ru were measured on the fuel rods, and consequently compared with the results of the ORIGEN2\\/82 calculation, in which both the original library and ORLIBJ32 based on the JENDL-3.2 library were used.

Akihiro SASAHARA; Tetsuo MATSUMURA; Giorgos NICOLAOU; Dimitri PAPAIOANNOU

2004-01-01

246

Thermochemical Prediction of Chemical Form Distributions of Fission Products in LWR Oxide Fuels Irradiated to High Burnup  

Microsoft Academic Search

Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial

Kouki MORIYAMA; Hirotaka FURUYA

1997-01-01

247

BWR Fuel Bundle Extended Burnup Program. Technical Progress Report, January 1982-December 1982.  

National Technical Information Service (NTIS)

At the start of this period there were four fuel bundles operating to extended burnup in the Monticello Nuclear Generating plant. All four of these bundles successfully completed their planned extended burnup operation and were discharged in September 198...

J. A. Baumgartner

1983-01-01

248

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

249

Isotopic biases for actinide-only burnup credit  

SciTech Connect

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

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

1997-04-01

250

Isotopic and criticality validation for actinide-only burnup credit  

SciTech Connect

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

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

1997-07-01

251

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

252

A validated methodology for evaluating burnup credit in spent fuel casks  

SciTech Connect

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

Brady, M.C. (Oak Ridge National Lab., TN (USA)); Sanders, T.L. (Sandia National Labs., Albuquerque, NM (USA))

1991-01-01

253

Dry Storage Demonstration for High Burnup Spent Nuclear Fuel-Feasibility Study.  

National Technical Information Service (NTIS)

Initially, casks for dry storage of spent fuel were licensed for assembly-average burnup of about 35 GWd/MTU. Over the last two decades, the discharge burnup of fuel has increased steadily and now exceeds 45 GWd/MTU. With spent fuel burnups approaching th...

M. A. McKinnon M. E. Cunningham

2003-01-01

254

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

255

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

PubMed

Correlations among the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), 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 (238)Pu/((239)Pu+(240)Pu) determined by alpha spectrometry, the alpha specific activities of Pu as well as the atom % abundances of the plutonium isotopes in the unknown samples were calculated without depending on mass spectrometry. The calculated alpha specific activities of Pu agreed with those determined by experiment within 2%, and the atom % abundances of the Pu isotopes agreed within 4% for (238)Pu, 5% for (239)Pu, 7% for (240)Pu and 5% for (242)Pu, respectively. In addition, an attempt was made to elucidate a correlation between the fuel burnup and the alpha activity ratio of (238)Pu/((239)Pu+(240)Pu) at a range of the burnup from 35.5 to 62.9 GWd/MtU. PMID:19914841

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

2009-11-01

256

Thermal Behavior of Advanced UO Fuel at High Burnup  

Microsoft Academic Search

To improve the fuel performance, advanced UO products are developed to reduce significantly Pellet-Cladding Interaction and Fission Gas Release to increase high burnup safety margins on Light Water Reactors. To achieve the expected improvements, doping elements are currently used, to produce large grain viscoplastic UO fuel microstructures. In that scope, AREVA NP is conducting the qualification of a new UO

E. Muller; T. Lambert; K. Silberstein; N. LHullier; C. Delafoy; B. Therache

2007-01-01

257

Benefits of actinide-only burnup credit for shutdown PWRs  

Microsoft Academic Search

Owners of PWRs that are shutdown prior to resolution of interim storage or permanent disposal issues have to make difficult decisions on what to do with their spent fuel. Maine Yankee is currently evaluating multiple options for spent fuel storage. Their spent fuel pool has 1,434 assemblies. In order to evaluate the value to a utility of actinide-only burnup credit,

D. Lancaster; E. Fuentes; C. Kang; D. Rivard

1998-01-01

258

The Fork+ burnup measurement system: Design and first measurement campaign  

Microsoft Academic Search

Previous work with the original Fork detector showed that burnup as determined by reactor records could be accurately allocated to spent nuclear fuel assemblies. The original Fork detector, designed by Los Alamos National Laboratory, used an ion chamber to measure gross gamma count and a fission chamber to measure neutrons from an activation source, ²⁴⁴Cm. In its review of the

C. E. Olson; D. R. Bronowski; W. McMurtry; R. Ewing; R. Jordan; D. Rivard

1998-01-01

259

Evaluation of driver fuel performance in the Joyo MK-II core  

SciTech Connect

Through cycle 24, 275 assemblies with [approx]35,000 fuel pins were irradiated without any indication of breach. The integrity of Joyo MK-II core fuel assemblies up to the design burnup limit was confirmed. The analysis and evaluation based on a large body of PIE data showed that the performance of the MK-II driver fuel was excellent and that lifetime could be expanded beyond the current burnup limit.

Katsuragawa, M.; Shikakura, S.; Iwanaga, S.; Nomura, S.; Shibahara, I.; Asaga, T. (Power Reactor and Nuclear Fuels Development Corp., Ibaraki-ken (Japan))

1992-01-01

260

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

261

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.  

National Technical Information Service (NTIS)

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 l...

T. A. Parish

1995-01-01

262

Analysis of Reactivity Measurements in Core Physics Experiments on Full-MOX BWR  

Microsoft Academic Search

A core physics experimental program FUBILA has been performed to study core physics characteristics of full-MOX BWR cores consisting of high Pu-enriched MOX assemblies for high burnups. The program includes the measurement of reactivity worth, which is essential in operating BWR cores. The reactivity worth is due to the reactivity caused by (1) changes in the in-channel void fraction of

Toru YAMAMOTO; Yoshihira ANDO; Tomohiro SAKAI; Koichi SAKURADA; Takuya UMANO

2009-01-01

263

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

264

Dissolution of low burnup Fast Flux Test reactor fuel  

Microsoft Academic Search

The first Fast-Flux Test Facility reactor fuel (mixed (U,Pu)Oâ composition) has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 1997°C. Linear dissolution rates of 0.99 to 1.57 mm\\/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 95°C

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

1984-01-01

265

Design of an advanced fork system for assembly burnup measurement  

Microsoft Academic Search

An Advanced Fork System has been designed to add gamma-ray collimation and spectroscopy capability to the Fork measurement system, which has been used for burnup verification at pressurized water reactors (PWR). The Advanced Fork System measures the neutron and gamma-ray yields and the energy spectnum of gamma-rays from spent fuel assemblies. A cadmium-zinc-telluride (CZT) crystal permits the identification of the

R. I. Ewing; K. D. Seager

1996-01-01

266

1. Core-Quasiparticle Coupling Model Calculations as a Test of IBA Core Descriptions of the Even-Mass Hg Isotopes. 2. Decay of Mass-Separated sup 203 At.  

National Technical Information Service (NTIS)

The Core-Quasiparticle Coupling Model (CQCM) for odd-mass nuclei, which is based on dynamical field theory and the Bardeen-Cooper-Schrieffer (BCS) method, has been applied to two problems. In the first, a study of Pauli exchange effects for the odd partic...

P. B. Semmes

1985-01-01

267

The Fork+ burnup measurement system: Design and first measurement campaign  

SciTech Connect

Previous work with the original Fork detector showed that burnup as determined by reactor records could be accurately allocated to spent nuclear fuel assemblies. The original Fork detector, designed by Los Alamos National Laboratory, used an ion chamber to measure gross gamma count and a fission chamber to measure neutrons from an activation source, {sup 244}Cm. In its review of the draft Topical Report on Burnup Credit, the US Nuclear Regulatory Commission indicated it felt uncomfortable with a measurement system that depended on reactor records for calibration. The Fork+ system was developed at Sandia National Laboratories under the sponsorship of the Electric Power Research Institute with the aim of providing this independent measurement capability. The initial Fork+ prototype was used in a measurement campaign at the Maine Yankee reactor. The campaign confirmed the applicability of the sensor approach in the Fork+ system and the efficiency of the hand-portable Fork+ prototype in making fuel assembly measurements. It also indicated potential design modifications that will be necessary before the Fork+ can be used effectively on high-burnup spent fuel.

Olson, C.E.; Bronowski, D.R.; McMurtry, W. [Sandia National Labs. (United States); Ewing, R. [Electric Power Research Inst. (United States); Jordan, R.; Rivard, D. [Maine Yankee Atomic Power Co., Westboro, MA (United States)

1998-12-31

268

Assessment of the Accuracy of Shape-Consistent Relativistic Effective Core Potentials Using Multireference Spin-Orbit Configuration Interaction Singles and Doubles Calculations of the Ground and Low-Lying Excited States of U4+ and U5+  

NASA Astrophysics Data System (ADS)

Multireference spin-orbit configuration interaction calculations were used to determine the accuracy of 60-, 68-, and 78-electron shape-consistent relativistic effective core potentials (RECPs) for uranium V and VI ground and low-lying excited states. Both 5fn and (5f6d)n, (n = 1, 2) reference spaces were investigated using correlation-consistent double-? quality basis sets. Accuracy was assessed against gas-phase experimental spectra. The 68-electron RECP calculations yielded low relative and rms errors and predicted the empirical ordering of states most consistently.

Beck, Eric V.; Brozell, Scott R.; Blaudeau, Jean-Philippe; Burggraf, Larry W.; Pitzer, Russell M.

2009-07-01

269

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

SciTech Connect

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

Tulenko, J.S. (Univ. of Florida, Gainesville (United States))

1992-01-01

270

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor  

SciTech Connect

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

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

2007-07-01

271

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

SciTech Connect

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

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

2012-07-01

272

Modeling of WWER-440 fuel pin behavior at extended burn-up  

Microsoft Academic Search

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70MWd\\/kgU. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution

Moustafa S El-Koliel; Attya A Abou-Zaid; A. A El-Kafas

2004-01-01

273

Effect of Polynomial Expansion Order of Intranode Flux Treatment in Nodal SN Transport Calculation Code NSHEX for Large-Size Fast Power Reactor Core Analysis  

Microsoft Academic Search

The nodal discrete ordinates (SN) transport calculation code for three-dimensional hexagonal geometry NSHEX treats intranode flux distribution using a polynomial series and considers the angular dependence of flux by the SN method. For the improvement of calculation accuracy of NSHEX for practical use to large-size fast reactor plants, the maximum order of the polynomial series is extended from two to

Kazuteru SUGINO; Teruhiko KUGO

2011-01-01

274

Calculational methodology and associated uncertainties: Sensitivity and uncertainty analysis of reactor performance parameters  

SciTech Connect

This chapter considers the calculational methodology and associated uncertainties both for the design of large LMFBR's and the analysis of critical assemblies (fast critical experiments) as performed by several groups within the US. Discusses cross-section processing; calculational methodology for the design problem; core physics computations; design-oriented approximations; benchmark analyses; and determination of calculational corrections and associated uncertainties for a critical assembly. Presents a detailed analysis of the sources of calculational uncertainties for the critical assembly ZPR-6/7 to illustrate the quantitative assessment of calculational correction factors and uncertainties. Examines calculational uncertainties that arise from many different sources including intrinsic limitations of computational methods; design-oriented approximations related to reactor modeling; computational capability and code availability; economic limitations; and the skill of the reactor analyst. Emphasizes that the actual design uncertainties in most of the parameters, with the possible exception of burnup, are likely to be less than might be indicated by the results presented in this chapter because reactor designers routinely apply bias factors (usually derived from critical experiments) to their calculated results.

Kujawski, E.; Weisbin, C.R.

1982-01-01

275

Monte Carlo and theoretical calculations of the first four perturbation coefficients in the high temperature series expansion of the free energy for discrete and core-softened potential models  

NASA Astrophysics Data System (ADS)

The first four perturbation coefficients in the expansion of the Helmholtz free energy in power series of the inverse of the reduced temperature for a number of potential models with hard-sphere cores plus core-softened and discontinuous tails are obtained from Monte Carlo simulations. The potential models considered include square-well, double square-well, and square-shoulder plus square-well, with different potential parameters. These simulation data are used to evaluate the performance of a traditional macroscopic compressibility approximation (MCA) for the second order coefficient and a recent coupling parameter series expansion (CPSE) for the first four coefficients. Comprehensive comparison indicates the incapability of the MCA for the second order coefficient in most non-stringent situations, and significance of the CPSE in accurately calculating these four coefficients.

Zhou, Shiqi; Solana, J. R.

2013-06-01

276

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

277

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

278

Use of WIMS-ANL lumped fission product cross sections for burned core analysis with the MCNP Monte Carlo code.  

National Technical Information Service (NTIS)

Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code an...

N. A. Hanan

1998-01-01

279

Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks  

Microsoft Academic Search

It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples

T. L. Sanders; W. H. Lake

1989-01-01

280

Fission-gas release in fuel performing to extended burnups in Ontario Hydro nuclear generating stations.  

National Technical Information Service (NTIS)

The average discharge burnup of CANDU fuel is about 200 MWh/kgU. A significant number of 37-element bundles have achieved burnups in excess of 400 MWh/kgU. Some of these bundles have experienced failures related to their extended operation. To date, hot-c...

M. R. Floyd J. Novak P. T. Truant

1992-01-01

281

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

Microsoft Academic Search

The influence of a burnup increase of a spent nuclear fuel on a deep geological disposal system was evaluated in this study. First, the impact of a burnup increase on each aspect related to thermal and nuclear safety concerns was quantified. And then, the tunnel length, excavation volume, and the raw materials for a cast insert, copper, bentonite, and backfill

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

2007-01-01

282

Limited Burnup Credit in Criticality Safety Analysis: A Comparison of ISG-8 and Current International Practice.  

National Technical Information Service (NTIS)

This report has been prepared to qualitatively assess the amount of burnup credit (reactivity margin) provided by ISG-8 compared to that provided by the burnup credit methodology developed and currently applied in France. For the purposes of this study, t...

I. C. Gauld

2001-01-01

283

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

Microsoft Academic Search

A fuel failure with an axial crack starting outside the cladding and penetrating inwards was experienced by high burnup BWR fuel rods in power ramp test. On the other hand, no fuel failure caused by power ramp test has been currently reported on PWR fuel rods at burnups higher than 50 GWd\\/t. Extensive research programs regarding hydrogen behaviors and mechanical

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

2007-01-01

284

Microstructural characterization of high burn-up mixed oxide fast reactor fuel  

NASA Astrophysics Data System (ADS)

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.

Teague, Melissa; Gorman, Brian; King, Jeffrey; Porter, Douglas; Hayes, Steven

2013-10-01

285

Structure of high-burnup-fuel Zircaloy cladding. [PWR; BWR  

SciTech Connect

Zircaloy cladding from high-burnup (> 20 MWd/kg U) fuel rods in light-water reactors is characterized by a high density of irradiation-induced defects (RID), compositional changes (e.g., oxygen and hydrogen uptake) associated with in-service corrosion, and geometrical changes produced by creepdown, bowing, and irradiation-induced growth. During a reactor power transient, the cladding is subject to localized stress imposed by thermal expansion of the cracked fuel pellets and to mechanical constraints imposed by pellet-cladding friction. As part of a program to provide a better understanding of brittle-type failure of Zircaloy fuel cladding by pellet-cladding interaction (PCI) phenomenon, the stress-rupture properties and microstructural characteristics of high-burnup spent fuel cladding have been under investigation. This paper reports the results of the microstructural examinations by optical microscopy, scanning (SEM), 100-keV transmission (TEM), and 1 MeV high-voltage (HVEM) electron microscopies of the fractured spent fuel cladding with a specific empahsis on a correlation of the structural characteristics with the fracture behavior.

Chung, H.M.

1983-06-01

286

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

287

The burnup dependence of light water reactor spent fuel oxidation  

SciTech Connect

Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5).

Hanson, B.D.

1998-07-01

288

Valence fluctuations in thin films and the ? and ? phases of Pu metal determined by 4f core-level photoemission calculations  

NASA Astrophysics Data System (ADS)

The average number of 5f electrons making up the valence state in plutonium metal together with the electronic fluctuations on each metal site has been a recent subject of debate. For the ? phase of Pu, where compared to the ? phase increased localization (more atomiclike character) leads to decreased overlap and volume increase, an f count close to either 5 or 6 has been proposed depending on the type of electronic structure calculation. In order to resolve the controversy, we analyze the Pu4f photoemission spectrum, which displays well screened and poorly screened peaks that can be used as a measure for the degree of localization. A simple analytical two-level model already shows on general grounds that the f count for Pu must be between 5 and 5.5. Furthermore, we present detailed Anderson impurity model calculations including the full multiplet structure for Pu4f photoemission, which are compared to previous experimental results obtained from 1 to 9 monolayers thin films of Pu on Mg and from Pu metal in the ? and ? phases. The trend in the satellite to main peak intensity ratio as a function of the Pu layer thickness gives a clear indication that Pu metal has an 5f5 like ground state. For the Pu allotropes and thicker films an f count of 5.22 is obtained with a Coulomb interaction U=4eV . The 5f fluctuations in Pu metal are very prominent and strongly material dependent. The calculations give a ground state with 9.6% f4 , 58.8% f5 , and 31.6% f6 for the ? phase and 5.7% f4 , 66.4% f5 , and 27.8% f6 for the ? phase while for the thin films the amount of f5 and the localization strongly increase with reduced thickness. The obtained findings are in agreement with recent electronic structure calculations for ? Pu using local-density approximation with dynamical mean-field theory and with the branching-ratio analysis of the PuN4,5 edge in electron-energy-loss spectroscopy.

van der Laan, G.; Taguchi, M.

2010-07-01

289

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

SciTech Connect

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

Cao, Y.; Gohar, Y.; Broeders, C. (Nuclear Engineering Division); (Inst. for Neutron Physics and Reactor Technology)

2010-10-01

290

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

291

Measurement and calculation of the fast-neutron and photon spectra from the core boundary to the biological shielding in the WWER-1000 reactor model.  

PubMed

The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised. PMID:16381689

Osmera, B; Cvachovec, F; Kyncl, J; Smutný, V

2005-01-01

292

The Encapsulated Nuclear Heat Source (ENHS) Reactor Core Design  

SciTech Connect

A once-for-life, uniform composition, blanket-free and fuel-shuffling-free reference core has been designed for the Encapsulated Nuclear Heat Source (ENHS) to provide the design goals of a nearly zero burnup reactivity swing throughout {approx}20 yr of full-power operation up to the peak discharge burnup of more than 100 GWd/t HM. What limits the core life is radiation damage to the HT-9 structural material. The temperature coefficients of reactivity are all negative, except for that of the coolant expansion. However, the negative reactivity coefficient associated with the radial expansion of the core structure can compensate for the coolant thermal expansion. The void coefficient is positive but of no safety concern because the boiling temperature of lead or lead-bismuth is so high that there is no conceivable mechanism for the introduction of significant void fraction into the core. The core reactivity coefficients, reactivity worth, and power distributions are almost constant throughout the core life.It was found possible to design such once-for-life cores using different qualities of Pu and transuranics as long as U is used as the primary fertile material. It is also feasible to design ENHS cores using nitride rather than metallic fuel. Relative to the reference metallic fuel core, nitride fuel cores offer up to {approx}25% higher discharge burnup and longer life, up to {approx}38% more energy per core, a significantly more negative Doppler reactivity coefficient, and less positive coolant expansion and coolant void reactivity coefficient but a somewhat smaller negative fuel expansion reactivity coefficient. The pitch-to-diameter ratio (1.45 of the nitride fuel cores using enriched N) is larger than that (1.36) for the reference metallic fuel core, implying a reduction of the coolant friction loss, thus enabling an increase in the power level that can be removed from the core by natural circulation cooling.It is also possible to design Pu-U(10Zr) fueled ENHS-type cores using Na as the primary coolant with either Na or Pb-Bi secondary coolants. The Na-cooled cores feature a tighter lattice and are therefore more compact but have spikier power distribution, more positive coolant temperature reactivity coefficients, and smaller reactivity worth of the control elements.

Hong, Ser Gi [Korea Atomic Energy Research Institute (Korea, Republic of); Greenspan, Ehud [University of California, Berkeley (United States); Kim, Yeong Il [Korea Atomic Energy Research Institute (Korea, Republic of)

2005-01-15

293

Core loading pattern optimization based on simulated annealing and successive linear programming  

Microsoft Academic Search

An algorithm is developed to determine optimum core loading patterns for a pressurized water reactor burnup cycle. The best locations of fuel assemblies are searched by simulated annealing. The optimum placement of burnable absorbers for any stochastically generated fuel loading pattern is determined using successive linear programming. Two-dimensional time-dependent problems are considered for numerical study.

Y. P. Mahlers

1995-01-01

294

Moderator poison design and burn-up calculations at the SNS  

Microsoft Academic Search

The spallation neutron source (SNS) at Oak Ridge National Laboratory was commissioned in April 2006. At the nominal operating power (1.4MW), it will have thermal neutron fluxes approximately an order of magnitude greater than any existing pulsed spallation source. It thus brings a serious challenge to the lifetime of the moderator poison sheets. The SNS moderators are integrated with the

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

2008-01-01

295

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

296

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

2011-01-01

297

Conservatism in the actinide-only burnup credit for PWR spent nuclear fuel packages  

SciTech Connect

In May 1995, the U.S. Department of Energy (DOE) submitted a topical report to the U.S. Nuclear Regulatory Commission (NRC) to gain actinide-only burnup credit for spent nuclear fuel (SNF) storage, transportation, or disposal packages. After approval of this topical report, DOE intends further submittals to the NRC to acquire additional burnup credit (e.g., the topical does not use fission products and is limited to only the first 100 yr of disposal). The NRC has responded to the topical with its preliminary questions. To aid in evaluation of the method, a review of the conservatism in the actinide-only burnup credit methodology was performed. An overview of the actinide-only burnup credit methodology is presented followed by a summary of the conservatism.

Lancaster, D.B. [TRW Environmental Safety Systems, Inc., Vienna, VA (United States); Rahimi, M. [Johnson & Associates, Inc., Vienna, VA (United States); Thornton, J. [Duke Engineering & Services, Vienna, VA (United States)

1996-12-31

298

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; Wagner, John C [ORNL; Mueller, Don [ORNL; Gauld, Ian C [ORNL

2011-01-01

299

Improved lumped model for thermal analysis of high burn-up nuclear fuel rods  

Microsoft Academic Search

High burn-up nuclear fuel elements have been intensively studied for prolonged lifetime of existing reactors and for next-generation advanced reactors. This paper presents an improved lumped-differential formulation for one-dimensional transient heat conduction in a heat generating cylinder with temperature-dependent thermo-physical properties typical of high burn-up nuclear fuel rods. Two-points Hermite approximations for integrals (H1,1\\/H1,1) were used to obtain the average

Auro C. Pontedeiro; Renato M. Cotta; Jian Su

2008-01-01

300

TEM analysis of pellet-cladding bonding layer in high burnup BWR fuel  

Microsoft Academic Search

Detailed analysis of the pellet-cladding bonding layer in high burnup nuclear fuel has been done by transmission electron microscopy (TEM). A specimen was prepared from the fuel, which had been irradiated to the pellet average burnup of 49 GWd\\/tU (1.2×1021 fissions\\/cm3) in a boiling water reactor (BWR). A 20 ?m thick bonding layer which consisted of two regions was observed.

K. Nogita; K. Une; Y. Korei

1996-01-01

301

Irradiation performance of metallic driver fuel in Experimental Breeder Reactor II to high burnup  

Microsoft Academic Search

The Experimental Breeder Reactor II Mark-II metallic-driver-fuel element has been irradiated to high burnup to assess element lifetime and performance reliability. The purpose of this paper is to describe the irradiation performance of the Mark-II fuel element to 10 at.% burnup. Fission gas behavior, fuel deformation, fuel-cladding chemical interaction, fuel-cladding mechanical interaction, and cladding dilation are examined for their effect

R. E. Einziger; B. R. Seidel

1980-01-01

302

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

303

Fission Gas Release Behavior from High Burnup UO2 Fuels under Rapid Heating Conditions  

Microsoft Academic Search

Fission gas release (FGR) behavior under rapid heating conditions of high burnup UO2 fuels with developed rim structure has been examined using two different out-of-pile heating techniques with no restraint pressure. The burnups of the fuel specimens were 36–86 GWd\\/tU. The bare fuel specimens were heated up to 600-1,800°C at heating rates of 1.7 to 4,600°C\\/s. The FGR process strongly

Katsumi UNE; Shinji KASHIBE; Akira TAKAGI

2006-01-01

304

Dissolution of low burnup Fast Flux Test reactor fuel  

SciTech Connect

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

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

1984-01-01

305

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

SciTech Connect

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

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

2012-07-01

306

Calculating Path-Dependent Travel Time Prediction Variance and Covariance for the SALSA3D Global Tomographic P-Velocity Model with a Distributed Parallel Multi-Core Computer  

NASA Astrophysics Data System (ADS)

Recently our combined SNL-LANL research team has succeeded in developing a global, seamless 3D tomographic P-velocity model (SALSA3D) that provides superior first P travel time predictions at both regional and teleseismic distances. However, given the variable data quality and uneven data sampling associated with this type of model, it is essential that there be a means to calculate high-quality estimates of the path-dependent variance and covariance associated with the predicted travel times of ray paths through the model. In this paper, we show a methodology for accomplishing this by exploiting the full model covariance matrix. Our model has on the order of 1/2 million nodes, so the challenge in calculating the covariance matrix is formidable: 0.9 TB storage for 1/2 of a symmetric matrix, necessitating an Out-Of-Core (OOC) blocked matrix solution technique. With our approach the tomography matrix (G which includes Tikhonov regularization terms) is multiplied by its transpose (GTG) and written in a blocked sub-matrix fashion. We employ a distributed parallel solution paradigm that solves for (GTG)-1 by assigning blocks to individual processing nodes for matrix decomposition update and scaling operations. We first find the Cholesky decomposition of GTG which is subsequently inverted. Next, we employ OOC matrix multiply methods to calculate the model covariance matrix from (GTG)-1 and an assumed data covariance matrix. Given the model covariance matrix we solve for the travel-time covariance associated with arbitrary ray-paths by integrating the model covariance along both ray paths. Setting the paths equal gives variance for that path. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

Hipp, J. R.; Encarnacao, A.; Ballard, S.; Young, C. J.; Phillips, W. S.; Begnaud, M. L.

2011-12-01

307

Irradiation performance of FBR Monju-type fuel with modified type 316 stainless steel at high burnup  

SciTech Connect

In the last three decades, extensive effort has been devoted to the development of austenitic stainless steels for the liquid-metal fast breeder reactor in Japan. In the driver fuels of the Joyo MK-II experimental fast reactor core, which has been operated successfully at 100 MW(thermal) since 1982, 127 fuel pins are assembled with wrapping wire into the hexagonal duct. The material for cladding, duct, and wire is 20% cold-worked modified Type 316 stainless steel (PNC316) and mixed-oxide (MOX) fuel pellets, with 29 wt% Pu content and 94% theoretical density (TD). The peak burnup and neutron dose at the current Joyo MK-II core has reached 84 GWd/tonne U and 50 dpa, respectively. The postirradiation examination of those assemblies showed excellent performance of the MOX fuel and a negligible pin diameter increase, which demonstrates that the conservative design for the Joyo MK-II driver fuels is satisfactory. A life-limiting factor of the austenitic steel fuel assemblies was demonstrated to be dominated by the swelling-induced bundle distortion in a duct. Specifically, for characterizing swelling of PNC316 materials, microstructure changes of the irradiated cladding and duct were extensively analyzed by means of transmission electron microscopy. In PNC316 the improvement of swelling resistance had been conducted by adjusting 20% cold working and by adding minor alloying elements such as P, B, Ti, and Nb within the specification range of chemical compositions in the standard Type 316 stainless steel.

Ukai, Shigeharu; Yoshitake, Ssunemitsu; Akasaka, Naoaki; Donomae, Takako; Katsuyama, Kozo; Mitsugi, Takeshi; Asaga, Takeo [PNC, Oarai (Japan)

1998-12-31

308

Pancake core high conversion light water reactor concept  

SciTech Connect

A new concept is proposed for a high conversion light water reactor (HCLWR) that achieves both high conversion and high burnup while maintaining a negative void reactivity coefficient. This HCLWR has a flat pancake core with thick axial blankets. By using the flat core, a potential problem of HCLWRs, the positive void reactivity coefficient can be reduced by neutron leakage, and a fuel assembly of very tight lattice pitch can be used. The leakage neutrons are utilized in the axial blankets to enhance the conversion ratio.

Ishiguro, Y.; Okumura, K.

1989-03-01

309

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

Microsoft Academic Search

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the

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

2010-01-01

310

Burnup verification measurements at U.S. Nuclear Facilities using the Fork system  

SciTech Connect

Burnup verification measurements have been performed using the Fork system at the Oconee Nuclear Station of Duke Power Company, and at Arkansas Nuclear One (Units 1 and 2), operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations ranged from 2.2% to 3.5% for assemblies at the three reactor sites, indicating a high degree of consistency in the reactor records. Anomalous measurements were also observed, but could be explained by the presence of neutron sources in the assemblies. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design.

Ewing, R.I.

1995-09-01

311

Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One  

SciTech Connect

Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The values of the exponent in the power laws were 3.83 and 4.35 for Units 1 and 2, respectively. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations were 2.7% and 3.5% for assemblies at Units 1 and 2, respectively, indicating a high degree of consistency in the reactor records. Two non-standard assemblies containing neutron sources were studied at Unit 2. No anomalous measurements were observed among the standard assemblies at either Unit. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design.

Ewing, R.I.

1995-09-01

312

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

313

Core design studies for a 1000 MW{sub th} advanced burner reactor.  

SciTech Connect

This paper describes the core design and performance characteristics of 1000 MW{sub th} Advanced Burner Reactor (ABR) core concepts with a wide range of TRU conversion ratio. Using ternary metal alloy and mixed oxide fuels, reference core designs of a medium TRU conversion ratio of {approx}0.7 were developed by trade-off between burnup reactivity loss and TRU conversion ratio. Based on these reference core concepts, TRU burner cores with low and high TRU conversion ratios were developed by changing the intra-assembly design parameters and core configurations. Reactor performance characteristics were evaluated in detail, including equilibrium cycle core performances, reactivity feedback coefficients, and shutdown margins. The results showed that by employing different assembly designs, a wide range of TRU conversion ratios from {approx}0.2 to break-even can be achieved within the same core without introducing significant performance and safety penalties.

Kim, T. K.; Yang, W. S.; Grandy, C.; Hill, R.; Nuclear Engineering Division

2009-04-01

314

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...

315

Dislocations in diamond: Core structures and energies  

Microsoft Academic Search

The structures and core energies of dislocations in diamond are calculated using both isotropic and anisotropic elasticity theory combined with ab initio-based tight-binding total energy calculations. Perfect and dissociated 60° and screw dislocations are considered. Their possible dissociation reactions are investigated through a consideration of the calculated elastic energy factors and core energies. Dissociation into partials is energetically favored. We

A. T. Blumenau; M. I. Heggie; C. J. Fall; R. Jones; T. Frauenheim

2002-01-01

316

The use of WIMS-ANL lumped fission product cross sections for burned core analysis with the MCNP Monte Carlo code  

Microsoft Academic Search

Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code and processed for use in MCNP. Results of analyses for four very different reactor cores using MTR-type and Russian-designed fuel

Hanan

1998-01-01

317

The use of WIMS-ANL lumped fission product cross sections for burned core analysis with the MCNP Monte Carlo code.  

SciTech Connect

Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code and processed for use in MCNP. Results of analyses for four very different reactor cores using MTR-type and Russian-designed fuel assemblies, with LEU and HEU fuels, are provided to demonstrate the use of this method.

Hanan, N. A.

1998-10-14

318

RTOP-Code Simulation of the Radial Distribution of Heat Release and Plutonium Isotope Accumulation in High Burnup Oxide Fuel  

Microsoft Academic Search

A model of radial profiles of burnup, heat release, and accumulation of plutonium isotopes is described. The model was developed for use in the mechanistic RTOP fuel element code. The model is based on theoretical ideas about the mechanisms leading to the formation of the radial burnup profile and a simplified description of the neutron spectrum in the reactor, employing

S. Yu. Kurchatov; V. V. Likhanskii; A. A. Sorokin; O. V. Khoruzhii

2002-01-01

319

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

320

Impact of high-burnup fuel cycles on fuel storage systems  

SciTech Connect

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 typical high density pool storage, consolidated fuel storage, and dry storage. System capacity, shielding requirements, thermal hydraulics, criticality control, and cost are addressed, as applicable. Specific systems were selected for evaluation. The pool selected holds 1500 pressurized water reactor (PWR) storage sleeves in high-density racks.

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

1986-01-01

321

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

322

Qualification of the APOLLO2.8 code package for the calculation of the fuel inventory and reactivity loss of UO x spent fuels in BWRs  

Microsoft Academic Search

This paper describes the neutronics qualification of the APOLLO2.8 code package for the calculation of UOx fuel inventory and reactivity loss with burn-up in Boiling Water Reactors (BWR).In the first part, the study deals with the analyses of spent fuels through the isotopic content of actinides (U, Pu, Np, Am, Cm), fission products used to evaluate the burn-up (Nd, Cs)

Pierre Leconte; Jean-François Vidal; David Bernard; Alain Santamarina; Romain Eschbach; Jean-Pascal Hudelot

2009-01-01

323

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

324

Degraded core modeling in MELCOR  

SciTech Connect

A package of phenomenological models has been developed for the MELCOR code system to calculate the thermal response of structures in the core and lower plenum of an LWR during a severe accident. This package treats all important modes of heat transfer within the core, as well as oxidation, debris formation, and relocation of core and structural materials during melting, candling, and slumping. Comparison of MELCOR and MARCON calculations for the Browns Ferry BWR primary system shows many areas of agreement during the early stages of core heatup and oxidation, but very large differences at later times. Many of these differences are attributed to the effects of candling predicted by MELCOR and the lack of any mechanistic candling or debris relocation models in MARCON. The melting and slumping behavior calculated by MELCOR is in qualitative agreement with our current understanding of the processes involved.

Summers, R.M.

1986-01-01

325

Burnup verification measurements at a US nuclear utility using the FORK measurement system  

SciTech Connect

The FORK measurement system, designed at Los Alamos National Laboratory (LANL) for the International Atomic Energy Agency (IAEA) safeguards program, has been used to examine spent reactor fuel assemblies at Duke Power Company`s Oconee Nuclear Station. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. These measurements can be correlated with burnup and cooling time, and can be used to verify the reactor site records. Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. By taking into account the reduced reactivity of spent fuel due to its burnup in the reactor, burnup credit results in more efficient and economic transport and storage. The objectives of these tests are to demonstrate the applicability of the FORK system to verify reactor records and to develop optimal procedures compatible with utility operations. The test program is a cooperative effort supported by Sandia National Laboratories, the Electric Power Research Institute (EPRI), Los Alamos National Laboratory, and the Duke Power Company.

Ewing, R.I. [Sandia National Labs., Albuquerque, NM (United States); Bosler, G.E. [Los Alamos National Lab., NM (United States); Walden, G. [Duke Power Co., Charlotte, NC (United States)

1993-08-01

326

Isotopic validation for PWR actinide-only burnup credit using Mihama-3 data.  

National Technical Information Service (NTIS)

This report augments a US Department of Energy Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages submitted to the Nuclear Regulatory Commission (NRC). The purpose of the topical report was to obtain NRC approval on a generi...

M. Rahimi D. Lancaster

1996-01-01

327

Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks  

Microsoft Academic Search

This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures

W. H. Lake; T. L. Sanders; C. V. Parks

1990-01-01

328

Fission Gas and Iodine Release Measured Up to 15 GWd/T UO sub 2 Burnup.  

National Technical Information Service (NTIS)

A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO sub 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m...

A. D. Appelhans

1983-01-01

329

Extended Fuel-Burnup Demonstration Program. Semi-Annual Technical Progress Report, July 1981-December 1981.  

National Technical Information Service (NTIS)

The feasibility of increasing the region average discharge burnup of pressurized water reactor (PWR) fuel assemblies to values exceeding 40,000 MWD/MTU is studied. For the 3-loop 15 x 15 assembly plant, the loading arrangement for the second transition cy...

H. D. Moss

1982-01-01

330

Limits of burnup-dose loading of pyrocarbon-coated particles  

Microsoft Academic Search

Work performed under United States -German High Temperature Reactor ; Research Exchange Program. The limits of the burnup-dose loading of PyC coated ; (U,Th)Oâ fuel particles were previously investigated. The particles were ; heated to 1500 deg C without showing damage. Theoretical burnupdose loading ; limits were investigated. A comparison between the maximum tensile stress and ; the tensile strength

1973-01-01

331

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

SciTech Connect

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

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

1996-10-01

332

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

333

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY2011 Activities  

Microsoft Academic Search

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the

Michael A. Pope

2011-01-01

334

Evaluation of driver fuel performance in the Joyo MK-II core  

Microsoft Academic Search

Through cycle 24, 275 assemblies with [approx]35,000 fuel pins were irradiated without any indication of breach. The integrity of Joyo MK-II core fuel assemblies up to the design burnup limit was confirmed. The analysis and evaluation based on a large body of PIE data showed that the performance of the MK-II driver fuel was excellent and that lifetime could be

M. Katsuragawa; S. Shikakura; S. Iwanaga; S. Nomura; I. Shibahara; T. Asaga

1992-01-01

335

Electronic Calculators.  

National Technical Information Service (NTIS)

Because of the large number of different makes and models of electronic calculators available, it is often difficult to select the calculator best suited for a particular application. This compendium was written to aid potential calculator users in making...

E. A. Pfeiffer

1973-01-01

336

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

NASA Astrophysics Data System (ADS)

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

Assawaroongruengchot, Monchai

337

Core breaker  

SciTech Connect

This patent describes a continuous mining machine having at least a pair of horizontally spaced and aligned rotatable cutter drums, an improved core breaker operatively associated with the cutter drums and comprising: an arcuate fixed support disposed between the drums; an elongated bit holder fixed to and extending outwardly from the support; the holder structure having a series of longitudinally spaced, core breaker bit accommodating apertures extending therethrough; each aperture being defined by a smaller diameter outer bore, a larger diameter inner bore and a shoulder extending generally normal to the aperture axis separating the bores; and a bit element and expansion ring assembly disposed in operative association with the cutter drums, accommodated in each aperture.

Morgan, V.B.; Mc Kinney, J.F.

1987-06-02

338

Status of axial heterogeneous liquid-metal fast breeder reactor core design studies and research and development  

SciTech Connect

The current status of axial heterogeneous core (AHC) design development in Japan, which consists of an AHC core design in a pool-type demonstration fast breeder reactor (DFBR) and research and development activities supporting AHC core design, is presented. The DFBR core design objectives developed by The Japan Atomic Power Company include (a) favorable core seismic response, (b) core compactness, (c) high availability, and (d) lower fuel cycle cost. The AHC concept was selected as a reference pool-type DFBR core because it met these objectives more suitably than the homogeneous core (HOC). The AHC core layouts were optimized emphasizing the reduction of the burnup reactivity swing, peak fast fluence, and power peaking. The key performance parameters resulting from the AHC, such as flat axial power/flux distribution, lower peak fast fluence, lower burnup reactivity swing, etc., were evaluated in comparison with the HOC. The critical experiments at the Japan Atomic Energy Research Institute's Fast Critical Assembly facility demonstrate the key AHC performance characteristics. The large AHC engineering benchmark experiments using the zero-power plutonium reactor and the AHC fuel pin irradiation test program using the JOYO reactor are also presented.

Nakagawa, H.; Inagaki, T.; Yoshimi, H.; Shirakata, K.; Watari, Y.; Suzuki, M.; Inoue, K.

1988-11-01

339

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

340

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

SciTech Connect

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

David Petti; John Maki

2005-02-01

341

Core material types for sandwich structures  

NASA Astrophysics Data System (ADS)

Sandwich structures, which are composed of a thick core between two thin faces, are commonly used in many engineering applications because they combine high stiffness and strength with low weight. Depending on the application of a particular sandwich structure, various types of cores can be used. Cores with various two-dimensional patterns were studied in this thesis program. The mechanical characteristics of three types of core with isotropic patterns--triangular, hexagonal, and starcell--were studied as related to applications in sandwich structures. The Young's modulus, shear modulus, and Poisson's ratio were calculated for the three core types in the direction normal to the faces. The compressive buckling strength and shear buckling strength were calculated for the three core types by modeling each cell wall of a core as a plate under compressive or shear load. To verify this model, tests were conducted on scaled specimens to measure the compressive buckling strength of each core. The flexibilities of the three cores were also studied. Compliances for the three cores were measured using biaxial flexural tests. Tests were performed on each core types in which the deflection of a circular core sample loaded at its center was measured. The three isotropic core patterns exhibited distinct characteristics. In the direction normal to the faces, all three cores had the same stiffness. However, the triangular core had lower compressive and shear buckling strengths than the other two core types. The starcell core exhibited high flexibility compared to the other cores, indicating a potential for application in curved sandwich structures.

Kim, Beomkeun

1998-09-01

342

Luminosity calculation  

SciTech Connect

The luminosity of the Tevatron collider was calculated. The data used for the calculation are the flying wire transverse beam profile and the SBD bunch profile. For the 900 GeV/c mini beta runs, the calculation was compared to the CDF luminosity monitor. The ratio of the calculation and C:BOLUMP is 0.95. 1 ref., 9 figs., 4 tabs.

Hsueh, Shao-Yuan.

1989-12-01

343

Effects of stoichiometry on cladding reaction in mixed-oxide fuel at high burnup. [PuOâ--UOâ  

Microsoft Academic Search

The effects of initial stoichiometry on the character and extent of fuel-cladding chemical interaction (FCCI) for mixed-oxide fuels at O\\/M's of 1.984 and 1.966 irradiated in EBR-2 to peak burnup are described. The strong effect of initial O\\/M character and extent of FCCI persists to high burnup. Wastage correlations for mixed-oxide fuel based on data from higher O\\/M fuel must

L. A. Lawrence; J. W. Weber; J. L. Devary

1978-01-01

344

Passive nondestructive burnup monitoring of MNSR irradiated fuel by measuring photoneutrons produced within fission products.  

PubMed

A passive nondestructive method for monitoring of Syrian miniature neutron source reactor (MNSR) fuel burnup is introduced. The inner irradiation site design inside the Be reflector was exploited to measure the generated photoneutrons induced by fission products hard gamma radiation in the subcritical state. The photoneutron flux was measured using gold foils as a function of cooling time and operation power. For cooling time ranges between 10 and 25d, experiments show that (140)Ba is the extremely dominating inducer of photoneutrons and the measured flux is proportional to the accumulated (140)Ba. This result forms a new method base for MNSR fuel burnup monitoring. It might be used also as a safeguards technique to check the operator declared information. PMID:19620012

Haddad, Kh

2009-06-26

345

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

346

Prestellar Cores in the Coalsack  

NASA Astrophysics Data System (ADS)

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 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.; Rathborne, J.; Walsh, W.; Butner, H. M.

2011-09-01

347

Numerical simulation of mixed oxide fuel burnup taking account of double heterogeneity  

Microsoft Academic Search

The results of a numerical simulation of the burnup of mixed oxide fuel in VVER-1000 taking account of double heterogeneity\\u000a are presented. The fuel consists of two different materials: a depleted uranium matrix with small amounts of plutonium and\\u000a agglomerates consisting of particles with an elevated plutonium concentration. An important problem is taking double heterogeneity\\u000a into account correctly, making it

M. A. Kalugin; V. I. Kuznetsov

2009-01-01

348

SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT  

Microsoft Academic Search

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

Georgeta Radulescu; Don Mueller; John C Wagner

2009-01-01

349

Electrolysis of Burnup-Simulated Uranium Nitride Fuels in LiCl-KCl Eutectic Melts  

Microsoft Academic Search

The electrochemical behavior of burnup-simulated uranium nitride fuels containing representative solid fission product elements, UN+Mo (Mo = 2.84 wt%), UN+Pd (Pd = 4.6 wt%) and (U, Nd)N (NdN = 8.0 wt%), was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl3 in order to clarify the effects of fission products on the dissolution of actinide nitrides and the

Takumi SATOH; Takashi IWAI; Yasuo ARAI

2009-01-01

350

Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks  

SciTech Connect

This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures are included, as well as a discussion of the economic and safety benefits of such designs.

Lake, W.H. (USDOE, Washington, DC (USA)); Sanders, T.L. (Sandia National Labs., Albuquerque, NM (USA)); Parks, C.V. (Oak Ridge National Lab., TN (USA))

1990-01-01

351

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

352

Modeling of the fuel burnup of research reactors on switching to low enrichment  

Microsoft Academic Search

The modeling of the neutronics of fuel assemblies on switching research reactors to low-enrichment fuel is examined. The specifications\\u000a and computational results for three test problems of modeling the isotopic composition of spent nuclear fuel from a research\\u000a reactor are presented. The test problems simulate fuel burnup in TVS-2M and -4M fuel assemblies of the IRT-1 reactor. It is\\u000a shown

V. G. Baranov; G. V. Tikhomirov; P. E. Kharitonov; A. V. Khlunov

2010-01-01

353

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

SciTech Connect

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

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

2002-02-15

354

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

SciTech Connect

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

Parks, C. V.

2000-03-13

355

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

356

IAEA sodium void reactivity benchmark calculations  

SciTech Connect

In this paper, the IAEA-1 992 ``Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core`` problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated.

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

1992-12-01

357

IAEA sodium void reactivity benchmark calculations  

SciTech Connect

In this paper, the IAEA-1 992 Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated.

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

1992-01-01

358

The influence of potassium on core and geodynamo evolution  

Microsoft Academic Search

We model the thermal evolution of the core and mantle using a parametrized convection scheme, and calculate the entropy available to drive the geodynamo as a function of time. The cooling of the core is controlled by the rate at which the mantle can remove heat. Rapid core cooling favours the operation of a geodynamo but creates an inner core

F. Nimmo; G. D. Price; J. Brodholt; D. Gubbins

2004-01-01

359

Calcium Calculator  

MedlinePLUS

... Fun Stuff Fun Stuff Best for Bones Food Calcium Calculator How much calcium do you get every day? Remember, you need ... Deal with Vitamin D List of Foods with Calcium & Vitamin D Calcium Calculator Fooling Around with Food ...

360

Metal fueled long life fast reactor cores with inherent safety features  

SciTech Connect

A large fast reactor core concept is proposed that has inherent safety characteristics against both the Unprotected Loss of Flow (ULOF) event and the Unprotected Transient of Over-Power (UTOP) event, where commonly used zirconium alloy metal fuel (U-Pu- Zr) is adopted to achieve a long life cycle length up to 5 years. The burn-up reactivity of the core which is equivalent to the maximum insertion reactivity in the UTOP due to the control rod run-out event at the rated power, is reduced to less than 1 $ by introducing minor actinides to the fuel, while the sodium void reactivity is suppressed to be negative by applying a step core concept, where the inner core height is lower than the outer core height, and by deleting the upper axial blanket. (authors)

Yokoyama, Tsugio [AITEL Corporation: 8-Shinsugita-cho, Isogo-ku Yokohama 235-8523 (Japan); Ninokata, Hisashi; Endo, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1 O-okayama, Meguro-ku, Tokyo, 152-8550 (Japan)

2007-07-01

361

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.

2009-01-15

362

Convection in the inner core  

NASA Astrophysics Data System (ADS)

Over the past two decades it has been debated if and why convection in the inner core might occur. Here we present a parameterized model for the thermal history of the inner core illustrating that convection is expected over a wide range of parameters. For a thermal conductivity of 36 W/Km, the critical CMB heat flux for convection is around 4 TW. A heat flux greater than 5.5 TW maintains convection in the inner core until the present day. In our model the growth of the inner core is calculated using a constant CMB heat flux. Latent and gravitational heat releases are incorporated. The temperature at the ICB is fixed by the liquidus. As the interior of the inner core cools, convection develops when the heat flux from the inner core exceeds the heat flux conducted down the adiabatic gradient. Fast growth during the early history of the inner core means that conduction alone is not sufficient to remove the heat. The convective heat flux is determined from the superadiabatic temperature gradient using a well-established relationship between the Rayleigh number (Ra) and the Nusselt number (Nu). Both the CMB heat flux and the physical parameters of iron at inner core conditions are poorly known, so we present a suite of solutions for a range of parameter values. The figure below shows the inner-core radius at which convection terminates for different choices of thermal conductivity and CMB heat flux. Convection is sustained for more of the inner-core history when thermal conduction is weak, or the growth rate is fast due to high CMB heat flux. The parameterized results are compared to numerical solutions using CitcomS. The numerical calculation illustrates how instabilities are formed close to the ICB and cause downwellings. Thermal convection can be suppressed prematurely (relative to the results shown in the figure) by a stable compositional gradient. Solidification excludes light elements from the inner core, causing the outer core to become more enriched in light elements. A stable compositional stratification develops when the composition of the crystallizing solid is proportion to the composition of the liquid. We predict the termination of convection when the stable compositional gradient suppresses the development of thermal instabilities at the inner-core boundary. Representative results suggest that convection is terminated for all but the highest values of CMB heat flow (for the above example, the critical heat flux for present day convection shifts from 5.5. TW to about 10 TW).

Cottaar, S.; Buffett, B. A.

2010-12-01

363

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

364

HOW STARLESS ARE STARLESS CORES?  

SciTech Connect

In this paper, we present the results of Combined Array for Research in Millimeter-wave Astronomy continuum and spectral line observations of the dense core Per-Bolo 45. Although this core has previously been classified as starless, we find evidence for an outflow and conclude that Per-Bolo 45 is actually an embedded, low-luminosity protostar. We discuss the impact of newly discovered, low-luminosity, embedded objects in the Perseus molecular cloud on starless core and protostar lifetimes. We estimate that the starless core lifetime has been overestimated by 4%-18% and the Class 0/I protostellar lifetime has been underestimated by 5%-20%. Given the relatively large systematic uncertainties involved in these calculations, variations on the order of 10% do not significantly change either core lifetimes or the expected protostellar luminosity function. Finally, we suggest that high-resolution (sub)millimeter surveys of known cores lacking near-infrared and mid-infrared emission are necessary to make an accurate census of starless cores.

Schnee, Scott; Friesen, Rachel [National Radio Astronomy Observatory, 520 Edgemont Road, Charlottesville, VA 22903 (United States); Di Francesco, James; Johnstone, Doug [National Research Council Canada, Herzberg Institute of Astrophysics, 5071 West Saanich Road Victoria, BC V9E 2E7 (Canada); Enoch, Melissa [Department of Astronomy, University of California, Berkeley, CA 94720 (United States); Sadavoy, Sarah, E-mail: sschnee@nrao.edu [Department of Physics and Astronomy, University of Victoria, Victoria, BC V8P 1A1 (Canada)

2012-01-20

365

Improved gas core propulsion model  

SciTech Connect

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

Tanner, J.E.

1993-10-01

366

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

National Technical Information Service (NTIS)

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron a...

A. Talamo Y. Gohar Z. Zhong

2009-01-01

367

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

NASA Astrophysics Data System (ADS)

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

Saigo, Kazuya; Tomisaka, Kohji; Matsumoto, Tomoaki

2008-02-01

368

Phasor Calculator  

NSDL National Science Digital Library

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

Reeder, Nick

2011-05-09

369

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

SciTech Connect

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

Primm, Trent [ORNL; Chandler, David [ORNL

2009-01-01

370

Core formation in giant gaseous protoplanets  

Microsoft Academic Search

Sedimentation rates of silicate grains in gas giant protoplanets formed by disk instability are calculated for protoplanetary masses between 1 MSaturn to 10 MJupiter. Giant protoplanets with masses of 5 MJupiter or larger are found to be too hot for grain sedimentation to form a silicate core. Smaller protoplanets are cold enough to allow grain settling and core formation. Grain

Ravit Helled; Gerald Schubert

2008-01-01

371

Constraints for a Solid Lunar Inner Core  

NASA Astrophysics Data System (ADS)

Whether or not the Moon has a core and whether or not that core would be fluid, solid or partly solid is still controversial. Recent reanalysis of Apollo seismic data [1] did not only determine the size of the lunar core more accurately but also suggested the presence of a solid inner and a fluid outer core. In order to constrain under which circumstances a solid inner core can form, we calculated a suite of three dimensional thermal evolution models of the Moon, where different parameters were varied. The results show that an inner core can indeed freeze out, assuming the amount of sulphur does not exceed 9 wt%. The observation of an inner core and its size can be used as an indirect constraint on the mantle rheology and composition.

Ziethe, R.; Spohn, T.

2012-09-01

372

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

373

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 Âtax supported entities world wide. 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 Âcalculator to include such things as name translators and databases on animal breeds.

374

Calcium Calculator  

MedlinePLUS

... form Search - A + A You are here Home - Calcium Calculator Printer friendly Email Share Tweet Like Home ... Regional Audits Reports Facts and Statistics Popular content Calcium-rich foods The Board What is Osteoporosis? Treating ...

375

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.

376

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

SciTech Connect

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

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

2007-07-01

377

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

SciTech Connect

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

Kimura, A.; Cho, H.S.; Lee, J.S.; Kasada, R. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Ukai, S. [Japan Nuclear Cycle Development Institute, Tokai (Japan); Fujiwara, M. [Kobelco, Ltd, Takatsukadai, Nishi-ku, Kobe (Japan)

2004-07-01

378

Laser Resonance Ionization Mass Spectrometry Measurements of Cesium in Nuclear Burn-up and Sediment Samples  

SciTech Connect

Isotopic ratio measurements of Cs-135 to Cs-137 Were performed using both resonance ionization mass spectrometry (RIMS) and thermal ionization mass spectrometry (TIMS) to determine the chronological age of nuclear fuel burn-up samples. Initial measurements on a lake sediment sample are being performed at NIST for determination of cesium content in the sample. Atomization behavior of the graphite furnace source, the overall efficiency and selectivity were measured for different sample preparations. Single-resonance excitation 6s S-2(1/2) (F = 4) --> 6p P-2(3/2) (F' = 5) with an extended cavity diode laser followed by photoionization with the 488 nm line of an argon ion laser yielded optical selectivity for Cs-135 and Cs-137 of more than two orders of magnitude against stable Cs-133 and overall selectivity of 10(8). An overall efficiency of 5 x 10(-7) was measured for standard Cs-133 solutions and for the nuclear fuel burn-up samples.

Pibida, L.; Mcmahon, C. A.; Bushaw, Bruce A.

2004-04-15

379

Core excitons in ionic crystals. III. Calculations (including correlation) of atomic energies required in the Born-Haber cycle for the Na LII,III exciton in NaF  

Microsoft Academic Search

For pt.II see ibid., vol.15, p.5037 (1982). The energies of key levels of low-lying configurations of the Na atom and the Na+ ion are calculated ab initio with inclusion of the effects of electron correlation. The levels considered are those required for the estimation (by the method described earlier in the present series of papers) of the energy of the

A. B. Kunz; T. O. Woodruff

1982-01-01

380

Banner Core Simulation Interface.  

National Technical Information Service (NTIS)

The Banner Core Simulation Interface tool has been developed by Rome Laboratory to support the Banner Core program in evaluating Bistatic Radar Systems. This document provides a summary of the objectives of the Banner Core program, the software capabiliti...

L. Mabius

2000-01-01

381

Biases for current FFTF calculational methods  

Microsoft Academic Search

Uncertainties in nuclear data and approximate calculational methods used in safety design, and operational support of a reactor yield biased as well as uncertain results. Experimentally based biases for use in Fast Flux Test Facility (FFTF) core calculations have been evaluated and are presented together with a description of calculational methods. Experimental data for these evaluations were obtained from an

P. A. Ombrellaro; R. A. Bennett; J. W. Daughtry; K. D. Dobbin; R. A. Harris; J. V. Nelson; R. E. Peterson; R. B. Rothrock

1978-01-01

382

Tracing Calculation (Calque Calcul) Between  

Microsoft Academic Search

To calculate the loss—is this the challenge that Nicolas Abraham has given to Jacques Derrida? Between 1959 and 1975, the year of Abraham's unexpected death, they were close friends, sharing what Elisabeth Roudinesco describes as \\

Nicolas Abraham; Jacques Derrida; Laurie Johnson

383

Finite element analysis of inductor core loss under DC bias condition  

Microsoft Academic Search

Finite element method is a popular way to analyze the magnetic core loss in complex core structures. But accurate calculation of the inductor core loss under DC bias condition is still a challenge because the magnetic properties like permeability and core loss density will change when a DC pre-magnetization is present, especially for saturable core. This paper proposes a method

Mingkai Mu; Feng Zheng; Qiang Li; Fred C. Lee

2012-01-01

384

Molecular evolution in star-forming cores: From prestellar cores to protostellar cores  

NASA Astrophysics Data System (ADS)

We investigate the molecular abundances in protostellar cores by solving the gas-grain chemical reaction network. As a physical model of the core, we adopt a result of one-dimensional radiation-hydrodynamics calculation, which follows the contraction of an initially hydrostatic prestellar core to form a protostellar core. Temporal variation of molecular abundances is solved in multiple infalling shells, which enable us to investigate the spatial distribution of molecules in the evolving core. The shells pass through the warm region of T ~ 20 100 K in several 104 yr and falls onto the central star in ~100 yr after they enter the region of T > 100 K. We found that the complex organic species such as HCOOCH3 are formed mainly via grain-surface reactions at T ~ 20 40 K, and then sublimated to the gas phase when the shell temperature reaches their sublimation temperatures (T ? 100 K). Carbon-chain species can be re-generated from sublimated CH4 via gas-phase and grain-surface reactions. HCO2+, which is recently detected towards L1527, are abundant at r = 100 2,000 AU, and its column density reaches ~1011 cm-2 in our model. If a core is isolated and irradiated directly by interstellar UV radiation, photo-dissociation of water ice produces OH, which reacts with CO to form CO2 efficiently. Complex species then become less abundant compared with the case of embedded core in ambient clouds. Although a circumstellar (protoplanetary) disk is not included in our core model, we can expect similar chemical reactions (i.e., production of large organic species, carbon-chains and HCO2+) to proceed in disk regions with T ~ 20 100 K.

Aikawa, Yuri; Wakelam, Valentine; Sakai, Nami; Garrod, R. T.; Herbst, E.; Yamamoto, Satoshi

2008-10-01

385

Buckling of Sandwich Composites; Effects of Core-Skin Debonding and Core Density  

NASA Astrophysics Data System (ADS)

Foam-core sandwich composites have been fabricated using innovative co-injection resin infusion technique and tested under in-plane compression. The sandwich construction consisted of Klegcell foam as core materials and S2-glass/vinyl ester composites as face sheets. Tests were conducted with various foam densities and also with implanted delamination between the core and the face sheet. The intent was to investigate the effect of core density, and the effect of core-skin debonds on the overall buckling behavior of the sandwich. Analytical and finite element calculations were also performed to augment the experimental observations. It has been observed that core density has direct influence on the global buckling of the sandwich panel, while embedded delamination seem to have minimal effect on both global as well as local buckling. Detailed description of the experimental work, finite element modeling and analytical calculations are presented in this paper.

Mahfuz, Hassan; Islam, Syful; Saha, Mrinal; Carlsson, Leif; Jeelani, Shaik

2005-03-01

386

Calculation Nation  

NSDL National Science Digital Library

These games from NCTM Illuminations are organized around content from the upper elementary and middle grades math curriculum. Students can play online math strategy games with opponents from anywhere in the world, or challenge themselves by investigating significant mathematical content and practicing fundamental skills. Games focus on fractions, factors, multiples, symmetry and more, as well as practice on basic multiplication and calculating area.

Illuminations, Nctm

2009-04-23

387

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.

Foundation, Wgbh E.

2010-12-23

388

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

389

Non-Destructive Fuel Burn-Up Study on WWR-SM Type Fuel Assemblies. Gamma Spectrometric Method.  

National Technical Information Service (NTIS)

A measuring method and an experimental facility consisting of a shielded sample container, a collimator tube and a Ge(Li) detector were developed to determine fuel burnup in WWR-SM type reactors. Axial and azimuthal isotope distributions were measured by ...

I. Vidovszky J. Vegh A. Kereszturi

1986-01-01

390

Fission Gas and Iodine Release Measured in IFA-430 Up to 15 GWd/T UO sub 2 Burnup.  

National Technical Information Service (NTIS)

The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Id...

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

1983-01-01

391

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

NASA Astrophysics Data System (ADS)

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.

2012-12-01

392

High Burnup Performance of an Advanced Oxide Fuel Assembly in FFTF (Fast Flux Test Facility) with Ferritic/Martensitic Materials.  

National Technical Information Service (NTIS)

An advanced oxide fuel assembly with ferritic/martensitic materials has successfully completed its sixth cycle of irradiation in the FFTF, reaching a peak pellet burnup greater than 100 MWd/KgM and a peak fast fluence greater than 15 x 10 sup 22 n/cm sup ...

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

1986-01-01

393

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

Microsoft Academic Search

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

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

2011-01-01

394

Comparisons of Neutron Cross Sections and Isotopic Composition Calculations for Fission-Product Evaluations  

Microsoft Academic Search

The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI (Korea Atomic Energy Research Institute)-BNL (Brookhaven National Laboratory) international collaboration have been compared with ENDF\\/B-VI.7. Also, the influence of the new evaluations on the isotopic composition calculations of the fission products has been estimated through the OECD\\/NEA burnup credit criticality benchmarks (Phase 1B) and

Do Heon Kim; Choong-Sup Gil; Jonghwa Chang; Yong-Deok Lee

2005-01-01

395

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

SciTech Connect

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

Ogata, Keizo; Aomi, Masaki; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization, 3-17-1 Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Etoh, Yoshinori [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Ito, Kunio [Grobal Nuclear Fuel - Japan Co., Ltd., 3-1 Uchikawa 2-chone, Yokosuka 239-0836 (Japan); Kido, Toshiya [Nuclear Development Corporation, 622-12 Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Teshima, Hideyuki [Mitsubishi Heavy Industries, Ltd. 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan)

2007-07-01

396

Core excitation effects in the breakup of halo nuclei  

NASA Astrophysics Data System (ADS)

The role of core excitation in the structure and dynamics of two-body halo nuclei is investigated. We present calculations for the resonant breakup of 11Be on protons at an incident energy of 63.7 MeV/nucleon, where core excitation effects were shown to be important. To describe the reaction, we use a recently developed extension of the DWBA formalism which incorporates these core excitation effects within the no-recoil approximation. The validity of the no-recoil approximation is also examined by comparing with DWBA calculations which take into account core recoil. In addition, calculations with two different continuum representations are presented and compared.

Moro, A. M.; de Diego, R.; Lay, J. A.; Crespo, R.; Johnson, R. C.; Arias, J. M.; Gómez-Camacho, J.

2012-10-01

397

Core excitation effects in the breakup of halo nuclei  

SciTech Connect

The role of core excitation in the structure and dynamics of two-body halo nuclei is investigated. We present calculations for the resonant breakup of {sup 11}Be on protons at an incident energy of 63.7 MeV/nucleon, where core excitation effects were shown to be important. To describe the reaction, we use a recently developed extension of the DWBA formalism which incorporates these core excitation effects within the no-recoil approximation. The validity of the no-recoil approximation is also examined by comparing with DWBA calculations which take into account core recoil. In addition, calculations with two different continuum representations are presented and compared.

Moro, A. M.; Diego, R. de; Lay, J. A.; Crespo, R.; Johnson, R. C.; Arias, J. M.; Gomez-Camacho, J. [Departamento de FAMN, Universidad de Sevilla, Apartado 1065, E-41080 Sevilla (Spain); Centro de Fisica Nuclear, Universidade de Lisboa, Av. Prof. Gama Pinto 2, 1649-003 Lisboa (Portugal) and Departamento de Fisica, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Av. Prof. Cavaco Silva, Taguspark (Portugal); Physics Department, University of Surrey, Guildford Surrey, GU2 7XH (United Kingdom); Departamento de FAMN, Universidad de Sevilla, Apartado 1065, E-41080 Sevilla (Spain); Departamento de FAMN, Universidad de Sevilla, Apartado 1065, E-41080 Sevilla (Spain) and Centro Nacional de Aceleradores, Universidad de Sevilla/Junta de Andalucia, E-41092 Sevilla (Spain)

2012-10-20

398

Enhancement of REBUS-3/DIF3D for whole-core neutronic analysis of prismatic very high temperature reactor (VHTR).  

SciTech Connect

Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing asymmetric absorber rods), surface-dependent discontinuity factors based on nodal equivalence theory have been introduced into the nodal diffusion theory option of the DIF3D code (DIF3D-nodal) to improve modeling accuracy. Additionally, the discontinuity factors based on the Simplified Equivalence Theory (SET) have been incorporated as an alternative and may be employed for both the DIF3D-nodal and DIF3D-VARIANT (nodal transport) solution options. Two- and three-dimensional core calculations have been performed using the routines developed and modified in this work, along with cross sections generated from single fuel block and one-dimensional or two-dimensional fuel-reflector model. Generally, REBUS-3/DIF3D results for the core multiplication factor and power distribution are found to be in good agreement with reference results (generated with MCNP continuous energy calculations) particularly when discontinuity factors are applied. The DIF3D-VARIANT option was found to provide a more accurate solution in its diffusion approximation than the DIF3D-nodal option. Control rod worths can be estimated with acceptably small errors compared to MCNP results. However, estimation of the core power tilt needs to be improved by introducing the surface-dependent discontinuity factor capability in DIF3D-VARIANT.

Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division

2006-10-13

399

EXPERIMENTS AND ANALYSIS ON BERYLLIUM OXIDE MODERATED CORES CONTAINING HYDROGEN  

Microsoft Academic Search

The reactivities of a number of cores containing BeO and polyethylene, ; which simulated water flooding, were measured and calculated. Other ; determinations included core symmetrics, reactivity worths of rods and rod voids, ; and perturbations on assembly reactivities. The experimental arrangements and ; measurements, the calculational methods used in the analysis, and the results of ; the analyticalexperimental comparison

W. B. Wright; R. G. Bardes

1962-01-01

400

Dynamic Desaturation Process in Saturated Cores Fault Current Limiters  

Microsoft Academic Search

The process of driving the magnetic core of saturated-cores fault-current-limiters (SCFCL) out of saturation has been studied numerically using finite elements calculations. The de-saturated core section was found to grow dynamically starting from the central region below the AC coils and gradually expanding towards the coil edges with increasing AC current amplitude. During the transition, the core volume underneath the

Y. Nikulshin; A. Friedman; Y. Wolfus; V. Rozenshtein; Y. Yeshurun

2012-01-01

401

Fission-detector determination of D-D triton burnup fraction in beam-heated TFTR (Tokamak Fusion Test Reactor) plasmas  

SciTech Connect

After the end of a neutral-beam injection pulse into a low-density TFTR plasma, once the beam-injected deuterons have thermalized, the neutron emission is dominated by the 14-MeV neutron production from D-D triton burnup. Ordinary fission detectors can measure the 14-MeV emission rate, which can be extrapolated back in time to estimate the equilibrium triton burnup fraction. The fractional burnup determined by this method is in the range of 0.3 to 1.5% for TFTR discharges to date, and is consistent with classical confinement and slowing down. 10 refs., 3 figs.

Jassby, D.L.; Hendel, H.W.; Barnes, C.W.; Bosch, S.; Cecil, F.E.; McCune, D.C.; Nieschmidt, E.B.; Strachan, J.D.

1987-06-01

402

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

403

Core polarization corrections to magnetic dipole moments and transitions  

Microsoft Academic Search

Core polarization corrections to ground state magnetic dipole moments are calculated in first-order perturbation theory, with the use of Sussex matrix elements. These corrections are introduced in the particle-(hole-) core coupling model. It is shown that the need for effectivegs factors in the calculation of magnetic moments can be explained largely in terms of one-particle-one-hole polarization effects of the core.

G. Vanden Berghe; K. Heyde; M. Waroquier

1973-01-01

404

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.

2009-01-08

405

NFE Core Bibliographies.  

National Technical Information Service (NTIS)

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 ...

1981-01-01

406

BN-600 full MOX core benchmark analysis.  

SciTech Connect

As a follow-up of the BN-600 hybrid core benchmark, a full MOX core benchmark was performed within the framework of the IAEA co-ordinated research project. Discrepancies between the values of main reactivity coefficients obtained by the participants for the BN-600 full MOX core benchmark appear to be larger than those in the previous hybrid core benchmarks on traditional core configurations. This arises due to uncertainties in the proper modelling of the axial sodium plenum above the core. It was recognized that the sodium density coefficient strongly depends on the core model configuration of interest (hybrid core vs. fully MOX fuelled core with sodium plenum above the core) in conjunction with the calculation method (diffusion vs. transport theory). The effects of the discrepancies revealed between the participants results on the ULOF and UTOP transient behaviours of the BN-600 full MOX core were investigated in simplified transient analyses. Generally the diffusion approximation predicts more benign consequences for the ULOF accident but more hazardous ones for the UTOP accident when compared with the transport theory results. The heterogeneity effect does not have any significant effect on the simulation of the transient. The comparison of the transient analyses results concluded that the fuel Doppler coefficient and the sodium density coefficient are the two most important coefficients in understanding the ULOF transient behaviour. In particular, the uncertainty in evaluating the sodium density coefficient distribution has the largest impact on the description of reactor dynamics. This is because the maximum sodium temperature rise takes place at the top of the core and in the sodium plenum.

Kim, Y. I.; Hill, R. N.; Grimm, K.; Rimpault, G.; Newton, T.; Li, Z. H.; Rineiski, A.; Mohanakrishan, P.; Ishikawa, M.; Lee, K. B.; Danilytchev, A.; Stogov, V.; Nuclear Engineering Division; International Atomic Energy Agency; CEA /Cadarache; SERCO Assurance; China Inst. of Atomic Energy; Forschnungszentrum Karlsruhe; Indira Gandhi Centre for Atomic Research; Japan Nuclear Cycle Development Inst.; Korea Atomic Energy Research Inst.; Inst. of Physics and Power Engineering

2004-01-01

407

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

National Technical Information Service (NTIS)

This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay he...

I. C. Gauld J. C. Ryman

2001-01-01

408

Core Competence and Education.  

ERIC Educational Resources Information Center

|Outlines the concept of core competence and applies it to postcompulsory education in the United Kingdom. Adopts an educational perspective that suggests accreditation as the core competence of universities. This economic approach suggests that the market trend toward lifetime learning might best be met by institutions developing a core…

Holmes, Gary; Hooper, Nick

2000-01-01

409

CONTINUOUSLY PRODUCED HONEYCOMB CORES  

Microsoft Academic Search

Today mechanical requirements and weight targets demand a lightweight sandwich design in many non-aerospace application areas. The large production cost of honeycomb core materials often prevents their use in low cost sandwich constructions. Other sandwich core materials however provide lower mechanical properties. The currently employed concepts for the production of honeycomb cores are reviewed to evaluate their potential for a

Jochen Pflug; Bart Vangrimde; Ignaas Verpoest

410

Herschel Galactic Cold Cloud Core Analysis  

NASA Astrophysics Data System (ADS)

We have compiled a sample of 106 lesser-known cores from the Herschel Galactic Cold Cloud Cores Key Program (Juvela, M. et al. 2007). Based on the assumption, that these represent the crowd of the cold cores in the galaxy well, we have started a deep individual investigation, beginning with a ground-based follow-up and molecular line measurement at IRAM 30m telescope. We present the methods and calculated values of the most important parameters on a selected source: the G130.38+11.25 molecular cloud, which is part of the L1340.

Verebelyi, Erika; Pagani, Laurent

2013-03-01

411

Effects of stoichiometry on cladding attack in mixed-oxide fuels to approx. 3. 6 at% burnup  

Microsoft Academic Search

Effects of initial stoichiometry on the character and extent of fuel-cladding chemical interaction (FCCI) were established for mixed-oxide fuels irradiated to peak burnups of approximately 3.6 at.%. The present LMFBR design basis considers an initial reduction of 50 microns (2 mils) in cladding thickness to compensate for possible FCCI. This thickness loss represents a significant part of the total cladding

D. C. Hata; L. A. Lawrence; J. W. Weber

1977-01-01

412

Measurement of power and burn-up in irradiated nuclear reactor fuel by a non-destructive method  

Microsoft Academic Search

A simple non-descructive test for measuring power and burn-up in irradiated nuclear reactor fuel is described. The method involves scanning the fuel elements via a slit in shielding for the gamma activity of two specific fission products. A short half-life fission product gives a measure of power production in the fuel for the period immediately prior to discharge, whilst a

T D Owen

1963-01-01

413

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

Microsoft Academic Search

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

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

2004-01-01

414

Results of AVR fuel pebble irradiation at increased temperature and burn-up in the HFR Petten  

Microsoft Academic Search

The irradiation experiment HFR-EU1bis was performed by the European Commission's Joint Research Centre-Institute for Energy (JRC-IE) in the HFR Petten to test five spherical High Temperature Reactor (HTR) fuel pebbles of former German production with TRISO coated particles for their potential for very high temperature performance and high burn-up. The irradiation started on 9 September 2004 and was terminated on

Michael A. Fütterer; Gerard Berg; Alain Marmier; Enrique Toscano; Daniel Freis; Klaas Bakker; Sander de Groot

2008-01-01

415

Behavior of low-burnup metallic fuels for the integral fast reactor at elevated temperatures in ex-reactor tests  

Microsoft Academic Search

A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel\\/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing

Hanchung Tsai; Yung Y. Liu; Da-Yung Wang; J. M. Kramer

1991-01-01

416

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

Microsoft Academic Search

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130GWd\\/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor,

Tomoyuki Uwaba; Masahiro Ito; Koji Maeda

2011-01-01

417

IRRADIATION RESULTS OF AVR FUEL PEBBLES AT INCREASED TEMPERATURE AND BURN-UP IN THE HFR PETTEN  

Microsoft Academic Search

The irradiation experiment HFR-EU1bis was successfully performed by JRC Institute for Energy (JRC-IE) in the HFR Petten to test 5 spherical High Temperature Reactor (HTR) fuel pebbles of former German production with TRISO coated particles for their potential for very high temperature performance and high burn-up. The irradiation started on 9 September 2004 and was terminated on 18 October 2005

Michael A. Fütterer; Enrique Toscano

418

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

Microsoft Academic Search

An advanced oxide fuel assembly with ferritic\\/martensitic materials has successfully completed its sixth cycle of irradiation in the FFTF, reaching a peak pellet burnup greater than 100 MWd\\/KgM and a peak fast fluence greater than 15 x 10SS n\\/cmS. The cladding, wire-wrap, and duct material for the ACO-1 test assembly is the ferritic\\/martensitic alloy, HT9, which was chosen for use

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

1986-01-01

419

Ice Core Secrets  

NSDL National Science Digital Library

In this activity, students will explore the characteristics of ice and explain the influencing factors by using Internet connections to polar field experiences, making their own ice cores and taking a field trip for obtaining a local ice core. The students will practice scientific journaling to document their observations. They will assemble their findings, develop a poster of their ice core and explain their observations. The 'ice is ice' misconception will be dispelled. Students will explain what scientists learn from ice cores and define basic vocabulary associated with ice cores.

Kolb, Sandra

420

Mixing core material into the envelopes of red grants  

SciTech Connect

A discussion is presented of calculations of four core helium flashes in red giant stars. The starting point for these calculations is a point source explosion on the polar axis of a two-dimensional finite difference grid. The amount of residue of the core helium flash mixed into and above the hydrogen shell is calculated at four temperatures for the elements carbon, oxygen, neon, magnesium, silicon, and sulfur. 7 refs., 1 tab.

Deupree, R.G.

1986-01-01

421

EXPOSED LONG-LIFETIME FIRST CORE: A NEW MODEL OF FIRST CORES BASED ON RADIATION HYDRODYNAMICS  

SciTech Connect

A first adiabatic core is a transient object formed in the early phase of star formation. The observation of a first core is believed to be difficult because of its short lifetime and low luminosity. On the basis of radiation hydrodynamic simulations, we propose a novel theoretical model of first cores, the Exposed Long-lifetime First core (ELF). In the very low-mass molecular core, the first core evolves slowly and lives longer than 10,000 years because the accretion rate is considerably low. The evolution of ELFs is different from that of ordinary first cores because radiation cooling has a significant effect there. We also carry out a radiation-transfer calculation of dust-continuum emission from ELFs to predict their observational properties. ELFs have slightly fainter but similar spectral energy distributions to ordinary first cores in radio wavelengths, therefore they can be observed. Although the probabilities that such low-mass cores become gravitationally unstable and start to collapse are low, we still can expect that a considerable number of ELFs can be formed because there are many low-mass molecular cloud cores in star-forming regions that could be progenitors of ELFs.

Tomida, Kengo; Tomisaka, Kohji [Department of Astronomical Science, The Graduate University for Advanced Studies (SOKENDAI), Osawa, Mitaka, Tokyo 181-8588 (Japan); Machida, Masahiro N.; Saigo, Kazuya [National Astronomical Observatory of Japan, Osawa, Mitaka, Tokyo 181-8588 (Japan); Matsumoto, Tomoaki, E-mail: tomida@th.nao.ac.j, E-mail: tomisaka@th.nao.ac.j, E-mail: masahiro.machida@nao.ac.j, E-mail: saigo.kazuya@nao.ac.j, E-mail: matsu@hosei.ac.j [Faculty of Humanity and Environment, Hosei University, Fujimi, Chiyodaku, Tokyo 102-8160 (Japan)

2010-12-20

422

Parallel plasma fluid turbulence calculations  

Microsoft Academic Search

The study of plasma turbulence and transport is a complex problem of critical importance for fusion-relevant plasmas. To this day, the fluid treatment of plasma dynamics is the best approach to realistic physics at the high resolution required for certain experimentally relevant calculations. Core and edge turbulence in a magnetic fusion device have been modeled using state-of-the-art, nonlinear, three-dimensional, initial-value

J. N. Leboeuf; B. A. Carreras; L. A. Charlton; J. B. Drake; V. E. Lynch; D. E. Newman; K. L. Sidikman; D. A. Spong

1994-01-01

423

Basic criticality relations for gas core design  

SciTech Connect

Minimum critical fissile concentrations are calculated for U-233, U-235, Pu-239, and Am-242m mixed homogeneously with hydrogen at temperatures to 15,000K. Minimum critical masses of the same mixtures in a 1000 liter sphere are also calculated. It is shown that propellent efficiencies of a gas core fizzler engine using Am-242m as fuel would exceed those in a solid core engine as small as 1000L operating at 100 atmospheres pressure. The same would be true for Pu-239 and possibly U-233 at pressures of 1000 atm. or at larger volumes.

Tanner, J.E.

1992-05-22

424

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

425

A stratified layer of light elements at the top of the outer core  

Microsoft Academic Search

Earth's core is thought to have formed from sinking metal diapirs that segregated at mid-mantle conditions. Consequently, the core and mantle may not be in chemical equilibrium. Recent experiments suggest that at the pressures and temperatures of the core, lower mantle oxides and silicates may have an increased solubility in iron. Geodynamic calculations predict that if a core\\/mantle chemical reaction

W. F. McDonough; B. A. Buffett; V. F. Cormier; S. Cottaar; E. A. Day; S. Dou; S. W. French; J. C. Irving; A. Kavner; M. P. Panning; R. Parai; I. Rose

2010-01-01

426

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

427

Effect of Test Blanket Module on Triton Burn-up in DIII-D Tokamak  

NASA Astrophysics Data System (ADS)

Time resolved measurements of triton burnup on DIII-D tokamak have been performed using a newly restored and upgraded 14 MeV neutron emission monitor based on silicon surface barrier diode. Neutron and energetic ion relevant data have been analyzed for the dedicated ITER Test Blanket Module (TBM) mockup plasma experiments. During the TBM on period, no observable change was recorded by a Faraday type fast ion loss collector, but a decrease in D-D neutron yield was usually encountered with a drop in plasma density. With full current applied on TBM coils, coincident reduction in 14 MeV neutron counts was prominent, as well as the ratio of 14 MeV D-T neutrons to 2.45 MeV D-D neutrons. In some cases, these behaviors were accompanied by a slight increase in signal of ion cyclotron emission loops attached behind the plasma-facing graphite protection module. Deterioration in triton confinement was observed to be weaker with partial current applied to the TBM and did not occur at all for no TBM current.

Zhu, Y. B.; Heidbrink, W. W.; Schaffer, M. J.

2010-11-01

428

Processing of low-burnup LEU (low enriched uranium) silicide targets  

SciTech Connect

Currently much of the world's supply of /sup 99m/Tc for medical purposes is produced from ZZMo derived from the fissioning of high enriched uranium (HEU). Substitution of low enriched uranium (LEU) silicide fuel for the HEU alloy and aluminide fuels used in current target designs will allow equivalent ZZMo yields with no change in target geometries. In these studies, targets were irradiated to low burnup (10/sup /minus/5/%) to produce fission products and STZNp at concentrations conveniently measured by gamma spectroscopy. Processing was done by dissolution of LEU targets in acid or base followed by alumina column recovery ZZMo. Acid dissolution is more rapid, but precipitation of silica results in loss of ZZMo. Dissolution of U3Si2--Al targets in base requires more processing steps than the current process for UAl/sub x/--Al fuel. A two-step process of first dissolving the 6061Al cladding and fuel meat aluminum, and then dissolving the U3Si2 fuel particles, has the advantage of eliminating the aluminum from further processing. Loss of ZZMo during the aluminum dissolution is attributed to recoil of ZZMo out of the silicide particles during irradiation. A larger particle size would decrease this ZZMo loss. 6 refs., 4 figs., 2 tabs.

Kwok, J.D.; Vandegrift, G.F.; Matos, J.E.

1988-09-01

429

Calculating conditional core damage probabilities for nuclear power plant operations  

Microsoft Academic Search

A part of managing nuclear power plant operations is the control of plant risk over time as components are taken out of service or plant upsets are caused by initiating events. Unfortunately, measuring risk over time proves to be challenging, even with modern probabilistic risk analyses (PRAs) and PRA tools. In general, the process of measuring the operational risk would

Curtis L. Smith

1998-01-01

430

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

431

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

432

Method of monitoring subcritical reactivity during core refueling  

SciTech Connect

The Modified Source Multiplication (MSM) technique, was used to measure core subcritical reactivity and core component reactivity worth during a major refueling of the Fast Flux Test Facility (FFTF) core. Measured results were compared with calculated values as a means of verifying core component changouts during the refueling. The results demonstrate the accuracy expected from MSM measurements and show that the technique was very effective in monitoring core subcritical reactivity during a refueling in which the neutron sources changed significantly and were not uniformly distributed.

Ombrellaro, P.A.; Harris, R.A.

1982-05-01

433

A combined study of electromagnetic core-mantle coupling. Inner core rotation and rotating magnetoconvection in the outer core  

NASA Astrophysics Data System (ADS)

Three aspects of deep Earth geophysics are investigated in this dissertation: (1)heterogeneous, electromagnetic coupling between the core and mantle, (2)rotational dynamics of the inner core, and (3)experimental rotating magnetoconvection in liquid gallium. Analytical calculations are performed to determine whether spatially heterogeneous electromagnetic torques between the core and mantle can explain virtual geomagnetic pole (VGP) paths found in some paleomagnetic studies of polarity reversals. These calculations show that during a reversal, the VGP is attracted to low conductivity regions in the lower mantle and repelled from high conductivity regions. Further calculations are made in which the core is free to rotate with respect to the mantle in response to the heterogeneous electromagnetic torques. In these calculations, resistive torques are induced by the differential motion between the mantle and the core and prove to be far greater than the heterogeneous torques. Thus, it is concluded that heterogeneous core-mantle torques have little effect on the position of the VGP during polarity reversals. The rotation of the solid inner core is studied with analytical and numerical models. These models show that it is possible to explain the seismically-inferred prograde rotation of the inner core [Song and Richards, 1996] by electromagnetically coupling the inner core to the thermally-driven, outer core fluid that is within the tangent cylinder, the imaginary cylinder tangent to the inner core equator. In contrast, the anomalous rotation of the inner core is not explicable in terms of a cylindrical model of flow, inferred from the geomagnetic westward drift. The interaction of electromagnetic, gravitational and mechanical inner core torques is also numerically modelled. The highest estimates of the gravitational torque produce a locked inner core that remains fixed in azimuth at the bottom of the gravitational well. For the lower bound estimate of the gravitational torque, electromagnetic torques can dominate and the inner core travels in and out of successive gravitational wells. The effect of mechanical torques is shown to be small in all cases. In a study of rotating magnetoconvection, experiments are performed in a plane layer of liquid gallium that is heated from below and cooled from above. The rotation axis and the imposed magnetic field are both aligned parallel to laboratory gravity (in the vertical direction). These experiments model the convection dynamics in the metallic outer core fluid located within the tangent cylinder. In magnetoconvection experiments, the onset of convection is inhibited by vertical magnetic fields. In rotating convection experiments, convection is strongly suppressed at high rotation rates and the onset of convection is oscillatory. The simultaneous effect of magnetic field and rotation always suppress thermal convection in the range of parameter space studied. Oscillatory convection is detected in the rotating convection experiments. In all of the convection experiments, time-dependent temperature signals are recorded very close to the onset of convection and, therefore, a regime of stationary convection is not detected. Thus convection in liquid gallium is inferred to be chaotic or even turbulent just beyond the onset of convection.

Aurnou, Jonathan Michael

2000-05-01

434

Anisotropic Earth's Inner Core within a Dynamic Core Scenario  

NASA Astrophysics Data System (ADS)

Recent global expansion of seismic data motivated a number of seismological studies of the Earth's inner core (EIC). An increasingly complex structure and anisotropy in EIC have been proposed to explain seismic data. In the meantime, new hypotheses of dynamic mechanisms have been put forward to interpret seismological results. In this study, the nature of anisotropy in EIC has been re-investigated by using PKP(BC-DF) core-sensitive differential travel-times and Fe-bcc/-hcp elastic constants calculated from first-principles. A Modified Transversely Isotropic Model (MTIM) has been introduced to account for a dynamic picture of EIC (e.g., eastward drift of material and heat flux variations at the CMB), where different chemical compositions could be stabilized at the polar/equatorial regions. Hemispherical patterns and anisotropic behaviour of EIC have been ascribed to the presence of denser polar regions (Si-poor Fe alloys) and lighter equatorial zones (Si-rich Fe alloys). A conglomerate-like EIC structure containing different material domains is then needed to address the complex anisotropy behaviour of the solid part of the Earth's core. Results have been discussed using both the seismic data from South Sandwich Islands (SSI) recorded in Alaska and the more recently collected travel-time residuals from the northern hemisphere to Antarctica.

Mattesini, M.; Belonoshko, A. B.; Tkalcic, H.; Buforn, E.; Ahuja, R.

2011-12-01

435

Glissile dislocations with transient cores in silicon.  

PubMed

We report an unexpected characteristic of dislocation cores in silicon. Using first-principles calculations, we show that all of the stable core configurations for a nondissociated 60 degrees dislocation are sessile. The only glissile configuration, previously obtained by nucleation from surfaces, surprisingly corresponds to an unstable core. As a result, the 60 degrees dislocation motion is solely driven by stress, with no thermal activation. We predict that this original feature could be relevant in situations for which large stresses occur, such as mechanical deformation at room temperature. Our work also suggests that postmortem observations of stable dislocations could be misleading and that mobile unstable dislocation cores should be taken into account in theoretical investigations. PMID:19792584

Pizzagalli, Laurent; Godet, Julien; Brochard, Sandrine

2009-08-07

436

On the consistent definition of spin-orbit effects calculated by relativistic effective core potentials with one-electron spin-orbit operators: Comparison of spin-orbit effects for Tl, TlH, TlH3, PbH2, and PbH4  

NASA Astrophysics Data System (ADS)

The spin-orbit effects for Tl, TlH, TlH3, PbH2, and PbH4 are evaluated by two-component calculations using several relativistic effective core potentials (RECP) with one-electron spin-orbit operators. The used RECPs are shape-consistent RECPs derived by Wildman et al. [J. Chem. Phys. 107, 9975 (1997)] and three sets of energy-consistent (or adjusted) RECPs published by Schwerdtfeger et al. [Phys. Scr. 36, 453 (1987); J. Chem. Phys. 90, 762 (1989)], Küchle et al. [Mol. Phys. 74, 1245 (1991)], and Leininger et al. [Chem. Phys. 217, 19 (1997)]. The shape-consistent RECP results are in very good agreement with the Küchle et al. energy-consistent RECP results for all the molecules studied here and all-electron results for TlH. The RECPs of Schwerdtfeger et al. and Leininger et al. seem to provide qualitatively different spin-orbit effects. If one defines spin-free RECP as the potential average of the corresponding two-component RECP, all RECPs give very similar spin-orbit effects for all the cases. Most of the discrepancies of molecular spin-orbit effects among various RECPs reported in the literature may originate from different definitions of RECPs with or without a spin-orbit term and not from the inherent difference in spin-orbit operators.

Han, Young-Kyu; Bae, Cheolbeom; Lee, Yoon Sup

1999-05-01

437

Dynamic origin of vortex core switching in soft magnetic nanodots.  

PubMed

The magnetic vortex with in-plane curling magnetization and out-of-plane magnetization at the core is a unique ground state in nanoscale magnetic elements. This kind of magnetic vortex can be used, through its downward or upward core orientation, as a memory unit for information storage, and thus, controllable core switching deserves some special attention. Our analytical and micromagnetic calculations reveal that the origin of vortex core reversal is a gyrotropic field. This field is induced by vortex dynamic motion and is proportional to the velocity of the moving vortex. Our calculations elucidate the physical origin of the vortex core dynamic reversal, and, thereby, offer a key to effective manipulation of the vortex core orientation. PMID:18232915

Guslienko, Konstantin Yu; Lee, Ki-Suk; Kim, Sang-Koog

2008-01-16

438

Ab initio shell model with a core  

Microsoft Academic Search

The No Core Shell Model (NCSM) has been successful in describing the properties of light nuclei, A <= 16, starting from the fundamental interactions among the A nucleons, but it is currently difficult to extend the NCSM to heavier nuclei, because of the extremely large model spaces involved in the calculations. We present a new procedure, based on performing two

Bruce Barrett; Alexander Lisetskiy; Michael Kruse; Petr Navratil; Ionel Stetcu; James Vary

2011-01-01

439

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

Microsoft Academic Search

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

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

2009-01-01

440

Ice Core Investigations  

NSDL National Science Digital Library

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 pollution to create a meaningful science learning experience for students.

Krim, Jessica; Brody, Michael