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Reactivity estimation for source-driven systems using first-order perturbation theory.

Description: Applicability of the first-order perturbation (FOP) theory method to reactivity estimation for source-driven systems is examined in this paper. First, the formally exact point kinetics equations have been derived from the space-dependent kinetics equations and the kinetics parameters including the dynamic reactivity have been defined. For the dynamic reactivity, exact and first-order perturbation theory expressions for the reactivity change have been formulated for source-driven systems. It has been also shown that the external source perturbation itself does not change the reactivity if the initial {lambda}-mode adjoint flux is used as the weight function. Using two source-driven benchmark problems, the reactivity change has been estimated with the FOP theory method for various perturbations. By comparing the resulting reactivity changes with the exact dynamic reactivity changes determined from the space-dependent kinetics solutions, it has been shown that the accuracy of the FOP theory method for the accelerator-driven system (ADS) is reasonably good and comparable to that for the critical reactors. The adiabatic assumption has also been shown to be a good approximation for the ADS kinetics analyses.
Date: July 2, 2002
Creator: Kim, Y.; Yang, W. S.; Taiwo, T. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

MUSE-4 experiment measurements and analysis.

Description: This report presents a review of the activities performed by the five teams involved in the MUSE-4 experimental program. More details are provided on the contribution by ANL during the year 9/02 to 9/03. The ANL activity consisted both in direct participation in the experimental measurements and in the physics analysis of the experimental data, mainly for the reactivity level, adjoint flux and fission rate distributions and the analysis of dynamic measurements for reactivity determination techniques in subcritical systems. The results provided to complete the Benchmark organized by the OECD and the CEA on the experiment MUSE-4 are also presented. Deterministic calculations have been performed via the ERANOS code system in connection with JEF2.2, ENDF/B-V and ENDF/B-VI data files.
Date: January 13, 2004
Creator: Aliberti, G.; Imel, G. & Palmiotti, G.
Partner: UNT Libraries Government Documents Department

Prompt Neutron Lifetime for the NBSR Reactor

Description: In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.
Date: June 24, 2012
Creator: Hanson, A.L. & Diamond, D.
Partner: UNT Libraries Government Documents Department

New concept for deep-penetration transport calculations and two new forms of the neutron transport equation

Description: A new concept to solve radiation transport problems is developed, bypassing the solution of the Boltzmann equation. A distribution function psi is defined as the product of the conventional neutron flux and adjoint distributions. Two equations, one complex and linear, the other real and nonlinear, are derived for psi. A conservation law for psi is established and a physical interpretation given for psi as a flux distribution for a limited number of source particles which will necessarily contribute to the integral response of interest. The linear but complex form of the transport equation is solved analytically for a sample case of a pure absorber in slab geometry.
Date: December 1, 1976
Creator: Gerstl, S. A. W.
Partner: UNT Libraries Government Documents Department

Automated Monte Carlo biasing for photon-generated electrons near surfaces.

Description: This report describes efforts to automate the biasing of coupled electron-photon Monte Carlo particle transport calculations. The approach was based on weight-windows biasing. Weight-window settings were determined using adjoint-flux Monte Carlo calculations. A variety of algorithms were investigated for adaptivity of the Monte Carlo tallies. Tree data structures were used to investigate spatial partitioning. Functional-expansion tallies were used to investigate higher-order spatial representations.
Date: September 1, 2009
Creator: Franke, Brian Claude; Crawford, Martin James & Kensek, Ronald Patrick
Partner: UNT Libraries Government Documents Department

COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

Description: COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B3 or B1 zero-dimensional approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constants may be output in any of several standard formats including INL format, ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional (1-D) discrete-ordinate transport code, is incorporated into COMBINE7.1. As an option, the 167 fine-group constants generated by zero-dimensional COMBINE portion in the program can be used to calculate regionwise spectra in the 1-D ANISN portion, all internally to reflect the 1-D transport correction. ...
Date: September 1, 2011
Creator: Yoon, Woo Y. & Nigg, David W.
Partner: UNT Libraries Government Documents Department

Accounting for time dependent source variations in surveillance dosimetry analysis

Description: One of the difficulties encountered in the calculation of dosimetry reaction rates is how to account for the time dependent behavior of the core source during the irradiation period. Indeed, even the obtaining of this source time dependence in adequate detail is not a trivial task. The straightforward approach of performing a DOT4 or similar transport calculation for each new relative source distribution, although correct, might be prohibitively expensive and time consuming when the irradiation period spans one or more complete fuel cycles, as it normally does. An alternative approach exists in the generation of a set of adjoint fluxes using DOT4 in the adjoint mode. Equations necessary for using adjoint approach on the Arkansas-1 reactor are presented.
Date: January 1, 1983
Creator: Maerker, R.E. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Applications guide to the MORSE Monte Carlo code

Description: A practical guide for the implementation of the MORESE-CG Monte Carlo radiation transport computer code system is presented. The various versions of the MORSE code are compared and contrasted, and the many references dealing explicitly with the MORSE-CG code are reviewed. The treatment of angular scattering is discussed, and procedures for obtaining increased differentiality of results in terms of reaction types and nuclides from a multigroup Monte Carlo code are explained in terms of cross-section and geometry data manipulation. Examples of standard cross-section data input and output are shown. Many other features of the code system are also reviewed, including (1) the concept of primary and secondary particles, (2) fission neutron generation, (3) albedo data capability, (4) DOMINO coupling, (5) history file use for post-processing of results, (6) adjoint mode operation, (7) variance reduction, and (8) input/output. In addition, examples of the combinatorial geometry are given, and the new array of arrays geometry feature (MARS) and its three-dimensional plotting code (JUNEBUG) are presented. Realistic examples of user routines for source, estimation, path-length stretching, and cross-section data manipulation are given. A deatiled explanation of the coupling between the random walk and estimation procedure is given in terms of both code parameters and physical analogies. The operation of the code in the adjoint mode is covered extensively. The basic concepts of adjoint theory and dimensionality are discussed and examples of adjoint source and estimator user routines are given for all common situations. Adjoint source normalization is explained, a few sample problems are given, and the concept of obtaining forward differential results from adjoint calculations is covered. Finally, the documentation of the standard MORSE-CG sample problem package is reviewed and on-going and future work is discussed.
Date: August 1, 1985
Creator: Cramer, S.N.
Partner: UNT Libraries Government Documents Department

Depth-charge static and time-dependence perturbation/sensitivity system for nuclear reactor core analysis. [LMFBR]

Description: This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code block for both static and time-dependence perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Labortary. The DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analysis of realistic multidimensional reactor models.
Date: September 1, 1981
Creator: White, J.R.
Partner: UNT Libraries Government Documents Department

Adjoint transport calculations for sensitivity analysis of the Hiroshima air-over-ground environment

Description: A major effort within the US Dose Reassessment Program is aimed at recalculating the transport of initial nuclear radiation in an air-over-ground environment. This paper is the first report of results from adjoint calculations in the Hiroshima air-over-ground environment. The calculations use a Hiroshima/Nagasaki multi-element ground, ENDF/B-V nuclear data, one-dimensional ANISN flux weighting for neutron and gamma cross sections, a source obtained by two-dimensional hydrodynamic and three-dimensional transport calculations, and best-estimate atmospheric conditions from Japanese sources. 7 references, 2 figures.
Date: January 1, 1984
Creator: Broadhead, B.L.; Cacuci, D.G. & Pace, J.V. III
Partner: UNT Libraries Government Documents Department