JENDLドシメトリー・ファイルに収納された⁹³Nb(n,n')^93mNbおよび¹⁹⁹Hg(n,n')^199mHg反応の中性子断面積

抄録

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.