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 04:00:00 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 Scientific Lab., NM (USA)
• DOE Contract Number: W-7405-ENG-36
• OSTI ID: 6797305
• Report Number(s): LA-7483-MS
• Country of Publication: United States
• Language: English

Related Subjects

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