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The MORSE code : a multigroup neutron and gamma-ray Monte Carlo transport code

Author: E A StrakerP N StevensD C IrvingV R CainU.S. Atomic Energy Commission.All authors
Publisher: Oak Ridge, Tenn. : Oak Ridge National Laboratory, 1970.
Series: ORNL (Series), 4585.; UC (Series)., 80,, Reactor Technology.
Edition/Format:   Print book : National government publication : EnglishView all editions and formats
Summary:
The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems may be obtained in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry, as well as specialized one-dimensional geometry  Read more...
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Additional Physical Format: Online version:
MORSE code.
Oak Ridge, Tenn. : Oak Ridge National Laboratory, 1970
(OCoLC)852766955
Material Type: Government publication, National government publication
Document Type: Book
All Authors / Contributors: E A Straker; P N Stevens; D C Irving; V R Cain; U.S. Atomic Energy Commission.; Oak Ridge National Laboratory.; United States. Defense Atomic Support Agency.
OCLC Number: 793818999
Notes: "September 1970."
"Note: This work partially funded by Defense Atomic Support Agency under subtask PEO8001."
Operated by Union Carbide Corporation for the U.S. Atomic Energy Commission.
Description: 302 pages : illustrations ; 27 cm.
Series Title: ORNL (Series), 4585.; UC (Series)., 80,, Reactor Technology.
Other Titles: Multigroup Oak Ridge Stochastic Experiment code (M0RSE)
Multigroup neutron and gamma-ray monte carlo transport code
Neutron Physics Division
Technical Report Archive & Image Library (TRAIL)
Responsibility: E.A. Straker [and others].

Abstract:

The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems may be obtained in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry, as well as specialized one-dimensional geometry descriptions, may be used with an albedo option available at any material surface. A detailed discussion of the relationship between forward and adjoint flux and collision densities, as well as a detailed description of the treatment of the angle of scattering, is given in the appendices. Logical flow charts for each subroutine add to the understanding of the code.

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