The NJOY Nuclear Data Processing System, Volume 1:User's Manual

In several important evaluations, however, the evaluator has divided the fission
process into parts: MT19, direct fission (n,f); MT20, second-chance fission
(n,n’)f; MT21, third-chance fission (n,2n)f; and MT38, fourth-chance fission
(n,3n)f. The procedure makes possible a more accurate representation of the
high-energy portion of the fission spectrum when fission is induced by neutrons
with energies above 5 or 6 MeV.
235U 238U
For such evaluations (for example, , ,
239
Pu), the following input is recommended.
3 18 *TOTAL FISSION*/
3 19 *(N,F)*/
3 20 *(N,N)F*/
3 21 *(N,2N)F*/
3 38 *(N,3N)F*/
6 19 *(N,F)*/
6 20 *(N,N)F*/
6 21 *(N,2N)F*/
6 38 *(N,3N)F*/
Note that 6/18 is omitted. A subsequent module, such as DTFR or MATXSR, can add
the partial matrices to obtain the total fission matrix.
A final complication of fission is the existence of delayed neutrons from
fission. For those materials that contain delayed neutron data, the user
should request these.
3 455 *DELAYED NUBAR*/
5 455 *DELAYED CHI*/
A later module can add the delayed data to the prompt matrix in order to obtain
“steady-state” values for ~af and x.
The matrix representation of fission is very general, but the results are
bulky and expensive, especially for large group structures (for example, the
620-group dosimetry set). For such cases, the short-cut fission spectrum option
(MFD=5) can be used. The MTD-value selects the parameters as follows.
MTD Spectrum Incident
Requested from MF/MT Energy
18 or 452 5/18 1 MeV
19 or 456 5/19 1 MeV
455 5/455 2 MeV
The incident energy can be changed at line 407 of GROUPR.
If the evaluation includes photon production data, GROUPR will prepare a
neutron-to-photon transfer matrix for each reaction requested. These reactions
are identified in the material dictionary by the presence of MF=12 (photon
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