Evaluation procedure of neutron cross sections and covariance matrices for the
93Nb(n, n')
93mNb and
199Hg(n, n')
199mHg reactions is described. The
C/E values of
235U fission spectrum averaged cross sections are 0.99 for the
93Nb(n, n')
93mNb reaction and 0.86 for the
199Hg(n, n')
199mHg reaction, respectively. These data are compiled in JENDL Dosimetry File and its revised version.
The
93Nb(n, n')
93mNb reaction is useful for surveillance dosimetry of reactor pressure vessel. The
199Hg(n, n')
199mHg reaction is useful for fast neutron measurement at low neutron flux field such as critical assembly field. Especially, cross sections of
199Hg(n, n')
199mHg reaction are firstly introduced in an evaluated neutron data file.
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