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Title: Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

Abstract

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

Authors:
 [1];  [1];  [2];  [1];  [1];  [3];  [3];  [4];  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Univ. of Tennessee, Knoxville, TN (United States)
  3. Univ. of Michigan, Ann Arbor, MI (United States)
  4. Core Physics, Inc., Wilmington, NC (United States)
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Consortium for Advanced Simulation of LWRs (CASL)
Sponsoring Org.:
USDOE
OSTI Identifier:
1352789
Report Number(s):
ORNL/TM-2017/95
NT0304000; NEAF343; CASL-U-2017-1280-000
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
97 MATHEMATICS AND COMPUTING

Citation Formats

Kim, Kang Seog, Clarno, Kevin T., Gentry, Cole, Wiarda, Dorothea, Williams, Mark L., Kochunas, Brendan, Liu, Yuxuan, Palmtag, Scott, and Godfrey, Andrew T. Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR. United States: N. p., 2017. Web. doi:10.2172/1352789.
Kim, Kang Seog, Clarno, Kevin T., Gentry, Cole, Wiarda, Dorothea, Williams, Mark L., Kochunas, Brendan, Liu, Yuxuan, Palmtag, Scott, & Godfrey, Andrew T. Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR. United States. doi:10.2172/1352789.
Kim, Kang Seog, Clarno, Kevin T., Gentry, Cole, Wiarda, Dorothea, Williams, Mark L., Kochunas, Brendan, Liu, Yuxuan, Palmtag, Scott, and Godfrey, Andrew T. Wed . "Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR". United States. doi:10.2172/1352789. https://www.osti.gov/servlets/purl/1352789.
@article{osti_1352789,
title = {Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR},
author = {Kim, Kang Seog and Clarno, Kevin T. and Gentry, Cole and Wiarda, Dorothea and Williams, Mark L. and Kochunas, Brendan and Liu, Yuxuan and Palmtag, Scott and Godfrey, Andrew T.},
abstractNote = {The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.},
doi = {10.2172/1352789},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Wed Mar 01 00:00:00 EST 2017},
month = {Wed Mar 01 00:00:00 EST 2017}
}

Technical Report:

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  • 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
  • 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)more » 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.« less
  • 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
  • Two coupled neutron and photon multigroup cross-section libraries, derived from ENDF/B-V nuclear data, are described. The energy group structures, 61n/23..gamma.. and 22n/10..gamma.., are subsets of the Vitamin-E 174n/38..gamma.. group structure, and are tailored to the iron and sodium resonances, windows, and capture gamma-ray spectra. Each of the two libraries are available in two formats, the AMPX master format and the ANISN format. Cross sections for all materials in the Vitamin-E library were collapsed using a standard energy weighting function, and in addition, several cross-section sets for each of the major constituents of commercial grade sodium, stainless steel (types 304 andmore » 316), and carbon steel were derived using several problem-dependent weighting functions for averaging the fine groups. Effects of various group structures and weighting functions on the accuracy of the broad group libraries are studied by ANISN analysis of a typical sodium-iron shield configuration.« less
  • 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