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Standard thermal energy group structure for generation of thermal group constants from ENDF/B data

Description: The final specifications of a standard energy group structure for the generation of thermal group constants from ENDF/B data are presented. The report represents the work of a committee appointed by the Codes and Formats Subcommittee of the Cross Section Evaluation Working Group and is a parallel effort to work being done in the epithermal energy range. The thermal energy group structure specified in this report was accepted November 10, 1972, by the Cross Section Evaluation Working Group as the standard for generation of thermal group constants from ENDF/B data. The standard thermal group structure specified in this report is consistent with past design experience and thermal spectrum codes, and incorporates specific features for effects known to be important in nuclear design applications in the thermal energy range. Specific recommendations are made as to methods to be used for generation of thermal group constants. (auth)
Date: March 1, 1974
Creator: Finch, D.R.
Partner: UNT Libraries Government Documents Department

An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model. Quarterly Technical Progress Report, April 1 - June 30, 2000

Description: The status summary of Nuclear Energy Research Initiative (NERI) Tasks - Phase 1 are: Task 1--The development of the following methods in 1D slab geometry: (1) Homogenization and definition of discontinuity factors, (2) Group constants functionalization using assembly transport solution of multigroup eigenvalue problem with albedo boundary conditions, and (3) solving coarse-mesh effective few-group 1D QD moment equations using tables of data parameterized with respect to the ratio {rvec n} {center_dot} {rvec J}{sup G}/{tilde {phi}{sup G}} on boundaries. Status summary of NERI Tasks - Phase 1: Task 2--Development of a numerical method for solving the 2D few-group moment QD equations: (1) Development of a nodal discretization method for 2D moment QD equations, and (2) Development of an efficient iteration method for solving the system of equations of the nodal discretization method for 2D moment QD equations.
Date: July 25, 2000
Creator: Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S. & Smith, Kord S.
Partner: UNT Libraries Government Documents Department

An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model. Nuclear Energy Research Initiative (NERI) Program. Quarterly Technical Progress Report

Description: Status Summary of NERI Tasks - Phase 1 - Task 1. The development of the following methods in ID slab geometry: (1) Homogenization and definition of discontinuity factors, (2) Group constants functionalization using assembly transport solution of multigroup eigenvalue problem with albedo boundary conditions, and (3) Solving coarse-mesh effective few-group 1D QD moment equations using tables of data parameterized with respect to the ratio {rvec n} {center_dot} {bar J}{sup G}/{bar {phi}}{sup G} on boundaries. Status Summary of NERI Tasks - Phase 1 - Task 2. Development of a numerical method for solving the 2D few-group moment QD equations: (1) Development of a nodal discretization method for 2D moment QD equations, and (2) Development of an efficient iteration method for solving the system of equations of the nodal discretization method for 2D moment QD equations.
Date: April 28, 2000
Creator: Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S. & Smith, Kord S.
Partner: UNT Libraries Government Documents Department

Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model''

Description: OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations.
Date: August 4, 2003
Creator: Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S.; Smith, Kord S.; Clarno, Kevin; Hiruta, Hikaru et al.
Partner: UNT Libraries Government Documents Department

An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model. Nuclear Energy Research Initiative (NERI) Program. Quarterly Technical Progress Report, July 1 - September 30, 2000

Description: Status Summary of Nuclear Energy Research Initiative (NERI) Tasks: Task 1--The development of the following methods in ID slab geometry: (1) Homogenization and definition of discontinuity factors, (2) Group constants functionalization using assembly transport solution of multigroup eigenvalue problem with albedo boundary conditions, and (3) Solving coarse-mesh effective few-group 1D QD moment equations using tables of data parametrized with respect to the ratio {rvec n} {center_dot} {rvec {tilde J}{sup G}}/{tilde {phi}{sup G}} on boundaries. Task 2--Development of a numerical method for solving the 2D few-group moment QD equations: (1) Development of a nodal discretization method for 2D moment QD equations, and (2) Development of an efficient iteration method for solving the system of equations of the nodal discretization method for 2D moment QD equations. Task 3--The development of the following methods in 2D X-Y geometry: (1) homogenization and definition of discontinuity factors, (2) group constants functionalization using assembly transport solution of multigroup eigenvalue problem with albedo boundary conditions, and (3) solving coarse-mesh effective few-group QD moment equations using tables of data parametrized with respect to the ratio {rvec n} {center_dot} {rvec {tilde J}{sup G}}/{tilde {phi}{sup G}} on boundaries. Task 4--Development of a numerical method for solving the few-group moment QD equations in 3D geometry: (1) Development of a nodal method for discretization of 3D moment QD equations, and (2) Development of an efficient iteration method for solving the system of nodal discretized equations of moment QD equations in 3D geometry.
Date: October 31, 2000
Creator: Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S. & Smith, Kord S.
Partner: UNT Libraries Government Documents Department

Final Technical Report

Description: OAK B202 Final Technical Report. The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations.
Date: August 4, 2003
Creator: Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S.; Smith, Kord S.; Clarno, Kevin; Hiruta, Hikaru et al.
Partner: UNT Libraries Government Documents Department

Background cross section method as a general tool for reactor analysis

Description: The background cross section method (also called the self-shielding method) has been used extensively in fast reactor analysis. More recently it has also become important in thermal power reactor studies. This paper reviews current applications of the method and describes efforts underway at the Los Alamos Scientific Laboratory to improve the accuracy and reliability of the approach and to extend its applicability to graphite moderated systems and shielding problems. Improvements discussed include a method for automatically accounting for energy dependent buckling that resolves long-standing discrepancies in the calculation of iron reflected criticals and which promises to improve deep-penetration calculations in iron, methods for treating intermediate resonance effects, methods for treating double heterogeneity in gas-cooled reactors, the automatic calculation of Levine factors, improved treatments of elastic removal, and improvements in processing codes and formats.
Date: January 1, 1978
Creator: MacFarlane, R.E.; Kidman, R.B.; LaBauve, R.J. & Becker, M.
Partner: UNT Libraries Government Documents Department

LINX and BINX: CCCC utility codes for the MINX multigroup processing code

Description: The LINX and BINX codes were written to manipulate multigroup cross- section libraries in the CCCC format. The LINX code is used to merge two ISOTXS or BRKOXS libraries. The BINX code is used to convert any ISOTXS, BRKOXS, or DLAYXS library from binary to BCD mode or back with the option to list all or any part of the library. These codes are utilities for the MINX multigroup processing system. 2 tables (auth)
Date: January 1, 1976
Creator: MacFarlane, R.E. & Kidman, R.B.
Partner: UNT Libraries Government Documents Department

COVERX service module of the FORSS system. [LMFBR]

Description: The COVERX Service Module includes seven execution paths to aid in understanding and using multigroup cross-section covariance matrices contained in the standard interface file COVERX. The execution paths provide the following operations on COVERX file(s): list the contents of a COVERX file; allow adding new multigroup cross-section covariance matrices to an existing COVERX file; allow deletion of multigroup covariance matrices from an existing COVERX file; merge two COVERX files and creates a new file; change the mode of a file from unformatted to formatted and conversely; allow modification of the records contained in a COVERX file; and selectively edits or copies a file.
Date: April 1, 1980
Creator: Drischler, J.D.
Partner: UNT Libraries Government Documents Department

Evaluation and processing of nuclear data

Description: The role a nuclear data evaluator plays in obtaining evaluated nuclear data, needed for applications, from measured nuclear data is surveyed. Specific evaluation objectives, problems, and procedures are discussed. The use of nuclear systematics to complement nuclear experiment and theory is described. With the Evaluated Nuclear Data File (ENDF) as an example, the formatting, checking, and processing of nuclear data are discussed as well as the testing of evaluated nuclear data in the calculation of integral benchmark experiments. Other important topics such as the Probability Table Method and interrelation between differential and integral data are also discussed. 25 figures.
Date: January 1, 1980
Creator: Pearlstein, S.
Partner: UNT Libraries Government Documents Department

Comparative analysis of homogeneous and heterogeneous core critical experiments

Description: Comparative analyses of the homogeneous core critical assembly ZPPR-4 Phase 1 and the heterogeneous core critical assembly ZPPR-7 Phase A have been performed whereby the impact of identical data and data processing changes is assessed for both types of cores. The data changes reflect the differences between ENDF/B-III and ENDF/B-IV, and the changes in data processing reflect differences in the heterogeneity treatment and in the calculation of the elastic removal cross sections. These differences impact differently on the homogeneous and heterogeneous cores. Use of the ENDF/B-IV data and the more rigorous data processing techniques removes a large part of the C/E discrepancies noted in ZPPR-7 Phase A, and both types of cores are essentially predicted consistently.
Date: January 1, 1978
Creator: Kujawski, E.; Hartman, A.K. & Stewart, S.L.
Partner: UNT Libraries Government Documents Department

Sensitivity of the analysis of heterogeneous core critical assemblies to cell modelling

Description: The ZPPR-7 Phase A critical experiments have been analyzed using multigroup cross sections generated for several different cell models of the drawers and self-shielding approximations. Single drawer models and four-drawer models consisting of fuel and blanket regions are considered. The cross sections are self-shielded using both a ''multi-region equivalence relation'' and a ''two-region equivalence relation.'' Provided the multigroup constants are properly self-shielded, criticality and the reaction rates are only slightly sensitive to the cell modelling.
Date: January 1, 1978
Creator: Hartman, A.K.; Kujawski, E. & Stewart, S.L.
Partner: UNT Libraries Government Documents Department

VITAMIN-E: an ENDF/B-V multigroup cross-section library for LMFBR core and shield, LWR shield, dosimetry and fusion blanket technology

Description: The Department of Energy (DOE) Office of Fusion Energy (OFE) and the Division of Reactor Research and Technology (DRRT) jointly sponsored the development of a coupled fine-group cross section library. This 171-neutron, 36-gamma-ray group library was based upon ENDF/B-IV and was intended to be applicable to fusion reactor neutronics and LMFBR core and shield analysis. Versions of the library are available from the Radiation Shielding Information Center (RSIC) at the Oak Ridge National Laborary in both AMPX and CCCC formats. Computer codes for energy group collapsing, interpolation on Bondarenko factors for resonance self-shielding and temperture corrections, and various other useful data manipulations are also available. The experience gained in the generation, validation and utilization of this library along with its broad range of applicability has led to the request for updating this data set using ENDF/B-V. Additional support in this regard has been provided by the Defense Nuclear Agency (DNA) and by the Electric Power Research Institute (EPRI) in support of weapons analyses and light water reactor shielding and dosimetry problems, respectively. The purpose of the report is to provide detailed specifications and rationale for the proposed ENDF/B-V update (designated VITAMIN-E) to the VITAMIN-C library.
Date: February 1, 1979
Creator: Weisbin, C.R.; Roussin, R.W.; Wagschal, J.J.; White, J.E. & Wright, R.Q.
Partner: UNT Libraries Government Documents Department