Application of a library of processed ENDF/B-IV fission-product aggregate decay data in the calculation of decay-energy spectra. [FITPULS, in FORTRAN for CDC computers]

Technical Report · Fri Sep 01 00:00:00 EDT 1978

DOI:https://doi.org/10.2172/6797305· OSTI ID:6797305

LaBauve, R J; England, T R; George, D C; Stamatelatos, M G

Results from summation calculations by the CINDER-10 code and ENDF/B-IV decay, cross section, and yield data for fission pulses were incorporated into an ENDF/B-type format. The organization and content of this basic fine-group source-term library is described. In addition, two codes are described that provide pulse functions as fits to a user-specified multigrouping of the fine-group library. These can be readily used, essentially as Green's functions, to produce the spectra following any specific reactor power history. A particular set of fitted beta and gamma spectra having wide utility is described. Absorption effects are incorporated. 37 figures, 12 tables.

• Research Organization: Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 6797305
• Report Number(s): LA-7483-MS; TRN: 79-002499
• Resource Relation: Other Information: Portions of document are illegible
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
22 GENERAL STUDIES OF NUCLEAR REACTORS
COMPUTER CODES
F CODES
FISSION PRODUCTS
AFTER-HEAT
BETA-MINUS DECAY
BETA SPECTRA
CDC COMPUTERS
FORTRAN
GAMMA SPECTRA
MULTIGROUP THEORY
NUCLEAR DATA COLLECTIONS
REACTOR KINETICS
BETA DECAY
COMPUTERS
DECAY
ISOTOPES
KINETICS
NEUTRON TRANSPORT THEORY
PROGRAMMING LANGUAGES
RADIOACTIVE MATERIALS
SPECTRA
TRANSPORT THEORY
652026* - Nuclear Properties & Reactions
A=220 & above
Experimental- Spontaneous & Induced Fission- (-1987)
220100 - Nuclear Reactor Technology- Theory & Calculation