Data testing of ENDF/B-VI fission spectrum representations

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The treatment of the fission spectrum in transport calculations has undergone many refinements. The use of the same, incident neutron-independent fission spectrum for all materials in a single calculation was generally accepted. Later, it became evident that the fission spectrum is not only material dependent, but also depends on the incident neutron energy. Because the results of calculations (criticality or shielding) are sensitive to this effect, an increasing number of calculations are done with the full fission matrix. Another well-known refinement of the description of the fission process is the more detailed description of it as the sum of direct fission, first chance fission (E>~6 MeV), second chance fission (E>~12 MeV), and so on; in ENDF terminology, the fission process is differentiated as reactions MT=19, MT=20, MT=21, and MT=38 respectively, in addition to their sum, the total fission MT=18. While the evaluators of ENDF/B libraries generally made sure that the detailed fission cross sections (MF=3) sum to the total fission cross section, the fission spectrum (MF=5) of the total fission given in libraries is generally not equivalent to the effective fission spectrum obtained from all partial spectra.
Original languageEnglish
Title of host publicationTrans. Am. Nucl. Soc. (USA)
Place of PublicationUSA
PublisherAmerican Nuclear Society
Pages560 - 1
StatePublished - 1991

Bibliographical note

ENDF/B-VI fission spectrum representations;transport calculations;neutron-independent fission spectrum;criticality;shielding;full fission matrix;direct fission;first chance fission;second chance fission;total fission cross section;partial spectra;


  • fission
  • fission reactor theory and design
  • nuclear engineering computing
  • shielding


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