skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

Abstract

Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations.

Authors:
; ; ; ; ; ; ; ;
Publication Date:
Research Org.:
Oak Ridge National Lab., TN (USA)
Sponsoring Org.:
DOE/ER
OSTI Identifier:
6502303
Report Number(s):
ORNL-6618
ON: DE91000545; TRN: 91-000693
DOE Contract Number:
AC05-84OR21400
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
07 ISOTOPES AND RADIATION SOURCES; 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; NEUTRON REACTIONS; NUCLEAR DATA COLLECTIONS; NEUTRON SOURCES; DESIGN; PHOTONUCLEAR REACTIONS; BENCHMARKS; CALCULATION METHODS; COMPUTER ARCHITECTURE; CROSS SECTIONS; DATA BASE MANAGEMENT; LIBRARIES; BARYON REACTIONS; HADRON REACTIONS; MANAGEMENT; NUCLEAR REACTIONS; NUCLEON REACTIONS; PARTICLE SOURCES; RADIATION SOURCES; 070201* - Radiation Sources- Design, Fabrication & Operation; 651000 - Nuclear Physics

Citation Formats

Ford, W.E. III, Arwood, J.W., Greene, N.M., Moses, D.L., Petrie, L.M., Primm, R.T. III, Slater, C.O., Westfall, R.M., and Wright, R.Q. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies. United States: N. p., 1990. Web. doi:10.2172/6502303.
Ford, W.E. III, Arwood, J.W., Greene, N.M., Moses, D.L., Petrie, L.M., Primm, R.T. III, Slater, C.O., Westfall, R.M., & Wright, R.Q. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies. United States. doi:10.2172/6502303.
Ford, W.E. III, Arwood, J.W., Greene, N.M., Moses, D.L., Petrie, L.M., Primm, R.T. III, Slater, C.O., Westfall, R.M., and Wright, R.Q. 1990. "Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies". United States. doi:10.2172/6502303. https://www.osti.gov/servlets/purl/6502303.
@article{osti_6502303,
title = {Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies},
author = {Ford, W.E. III and Arwood, J.W. and Greene, N.M. and Moses, D.L. and Petrie, L.M. and Primm, R.T. III and Slater, C.O. and Westfall, R.M. and Wright, R.Q.},
abstractNote = {Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations.},
doi = {10.2172/6502303},
journal = {},
number = ,
volume = ,
place = {United States},
year = 1990,
month = 9
}

Technical Report:

Save / Share:
  • The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44gmore » coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).« less
  • Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory.
  • AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all writtenmore » in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.« less
  • A P/sub 3/ 227-neutron-group cross-section library has been processed for the subsequent generation of problem-dependent fine- or broad-group cross sections for a broad range of applications, including shipping cask calculations, general criticality safety analyses, and reactor core and shielding analyses. The energy group structure covers the range 10/sup -5/ eV - 20 MeV, including 79 thermal groups below 3 eV. The 129-material library includes processed data for all materials in the ENDF/B-V General Purpose File, several data sets prepared from LENDL data, hydrogen with water- and polyethyelene-bound thermal kernels, deuterium with C/sub 2/O-bound thermal kernels, carbon with a graphite thermalmore » kernel, a special 1/V data set, and a dose factor data set. The library, which is in AMPX master format, is designated CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data). Also included in CSRL-V is a pointwise total, fission, elastic scattering, and (n,..gamma..) cross-section library containing data sets for all ENDF/B-V resonance materials. Data in the pointwise library were processed with the infinite dilute approximation at a temperature of 296/sup 0/K.« less
  • The principal aim of this neutron cross-section research is to provide the utility industry with a 'standard nuclear data base' that will perform satisfactorily when used for analysis of thermal power reactor systems. EPRI is coordinating its activities with those of the Cross Section Evaluation Working Group (CSEWG), responsible for the development of the Evaluated Nuclear Data File-B (ENDF/B) library, in order to improve the performance of the ENDF/B library in thermal reactors and other applications of interest to the utility industry. Battelle-Northwest (BNW) was commissioned to process the ENDF/B Version-4 data files into a group-constant form for use inmore » the LASER and LEOPARD neutronics codes. Performance information on the library should provide the necessary feedback for improving the next version of the library, and a consistent data base is expected to be useful in intercomparing the versions of the LASER and LEOPARD codes presently being used by different utility groups. This report describes the BNW multi-group libraries and the procedures followed in their preparation and testing. (GRA)« less