ENDF reaction MT's and macroscopic reaction numbers

Serpent uses standard ENDF reaction MT's to identify neutron and photon reactions. The number are used with detector response functions and printed in various output files. Detector responses also include macroscopic cross sections (and similar), identified by negative reaction numbers.

Below are descriptive lists of ENDF reaction MT's and macroscopic reaction numbers. For more information on the MT numbers, see the ENDF Format Manual. It should be noted that even though the notation is very similar to that used by MCNP, there are some differences in the definition of some response functions.

ENDF Reaction MT's

MT	Description	Notes
1	total
2	elastic scattering
3	nonelastic
4	total inelastic scattering
5	anything	used for lumping together multiple reaction modes
11	(n,2nd)
16	(n,2n)
17	(n,3n)
18	total fission	sum over all fission channels (MT's 19-21 and 38)
19	(n,f)	1st-chance fission
20	(n,nf)	2nd-chance fission
21	(n,2nf)	3rd-chance fission
22	(n,nα)
23	(n,n3α)
24	(n,2nα)
25	(n,3nα)
27	absorption
28	(n,np)
29	(n,n2α)
30	(n,2n2α)
32	(n,nd)
33	(n,nt)
34	(n,n3He)
35	(n,nd2α)
36	(n,nt2α)
37	(n,4n)
38	(n,3nf)	4th-chance fission
41	(n,2np)
42	(n,3np)
44	(n,n2p)
45	(n,npα)
51-90	inelastic scattering to excited states
91	inelastic scattering to continuum
101	total absorption
102	(n,γ)
103	(n,p)
104	(n,d)
105	(n,t)
106	(n,3He)
107	(n,α)
108	(n,2α)
109	(n,3α)
111	(n,2p)
112	(n,pα)
113	(n,t2α)
114	(n,d2α)
115	(n,pd)
116	(n,pt)
117	(n,dα)
201	total neutron production	ote to developers: check if this needs to be multiplied by total xs
202	total photon production	Note to developers: check if this needs to be multiplied by total xs
203	total proton production	Note to developers: check if this needs to be multiplied by total xs
204	total deuteron production	Note to developers: check if this needs to be multiplied by total xs
205	total triton production	Note to developers: check if this needs to be multiplied by total xs
206	total 3He production	Note to developers: check if this needs to be multiplied by total xs
207	total α production	Note to developers: check if this needs to be multiplied by total xs
301	total heat production	Total heating number multiplied by total cross section (difference to MCNP)
443	kinematic KERMA	Note to developers: check if this needs to be multiplied by total xs
444	damage-energy production	Note to developers: check if this needs to be multiplied by total xs
600	(n,p) to ground state	MT's 600-649 can be used to replace MT 103
601-648	(n,p) to excited states
649	(n,p) to continuum
650	(n,d) to ground state	MT's 650-699 can be used to replace MT 104
651-698	(n,d) to excited states
699	(n,d) to continuum
700	(n,t) to ground state	MT's 700-749 can be used to replace MT 105
701-748	(n,t) to excited states
749	(n,t) to continuum
750	(n,3He) to ground state	MT's 750-799 can be used to replace MT 106
751-798	(n,3He) to excited states
799	(n,3He) to continuum
800	(n,α) to ground state	MT's 800-849 can be used to replace MT 107
801 - 848	(n,α) to excited states
849	(n,α) to continuum
875	(n,2n) to ground state	MT's 875-891 can be used to replace MT 16
876-890	(n,2n) to excited states
891	(n,2n) to continuum

Macroscopic reaction numbers

Reaction number	Description	Notes
-1	macroscopic total cross section
-2	macroscopic total absorption cross section
-3	macroscopic total elastic scattering cross secion
-6	macroscopic total fission cross section
-4	macroscopic total heating cross section
-5	macroscopic total photon production cross section
-7	macroscopic total fission neutron production cross section ($\nu\Sigma_\mathrm{f}$)		-8	macroscopic total fission energy production cross section ($\kappa\Sigma_\mathrm{f}$)		-9	majorant cross section
macroscopic scattering recoil energy production cross section
source rate
neutron density
macroscopic total scattering neutron production cross section
temperature majorant cross section