Is there a way to calculate six group of effective delayed neutron data (delayed neutron fractions, yields and decay constants ) by MCNP code? Also, Is there a way to calculate nine group of delayed photoneutrons data using MCNP code?
The TOTNU card calculate the value of effective multiplication factor for only prompt
neutrons. If the TOTNU card is absent or if a TOTNU card is used but has no entry after it, the total average number of neutrons from fission (ν) using both prompt and delayed neutrons is applied and the total effective multiplication factor (k) is calculated. We can calculate the total effective delayed neutron fraction using (k2-k1)/(k2k1)
where k2 and k1 are the reactor multiplication factors with and without
delayed neutrons.
I applied the TOTNU card in my previous research for calculation of total effective delayed neutron fraction "Conceptual design study of the low power and LEU medical reactor for BNCT using in-tank fission converter to increase epithermal flux. Progress in Nuclear Energy, 95, 70-77."
But i need to calculate six group of effective delayed neutron fraction.
Effective delayed neutron fraction, βeff can be directly calculated with the enhanced feature of MCNP5 for adjoint weighted calculation as in the following reference,
B.C. Kiedrowski, T.E. Booth, F.B. Brown, J.S. Bull, J.A. Favorite, R.A. Forster, R.L. Martz, “MCNP5-1.60 Feature Enhancement & Manual Clarifications”, Los Alamos National Laboratory, LA-UR-10-06217, (2010).