|
In order to follow the transport of nuclear radiation through a
material, it is important to know which secondary products are
produced, the yield of each product, and how each product is
distributed in energy and angle. The capabilities of the
ENDF-format to represent this information about the products
has evolved from fairly simple representations using Files 4 and
5 to the current rather complete capabilities found in File 6.
In general, the cross section (in barns/steradian) for producing
a particle can be written
in terms of a cross section σ(E), a yield y(E), and a normalized distribution in initial energy E, final energy E', and cosine μ. The cross section is always given in File 3. The yield may be implicit as determined by the MT number, or in File 6, it may be given explicitly as integers for simple reactions or in noninteger form for the complex summation reaction MT=5. The distributions may be represented using three different approaches:
The File 5 representation is always used for fission in ENDF files. The neutrons are assumed to be emitted isotropically in the laboratory reference frame. |
NEXT | INDEX | ||
15 December 2012 | T-2 Nuclear Information Service | ryxm@lanl.gov |