ENDF reaction MT's and macroscopic reaction numbers
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Serpent uses standard ENDF reaction MT's to identify neutron and photon reactions. The numbers 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.[1] 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.
Contents
ENDF Reaction MT's
Neutron reactions
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 |
Photon reactions
Macroscopic reaction numbers
Neutron reactions
Reaction # | 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 | equivalent with the F8 tally in MCNP |
-5 | macroscopic total photon production cross section | |
-7 | macroscopic total fission neutron production cross section | |
-8 | macroscopic total fission energy production cross section | |
-9 | majorant cross section | |
-10 | macroscopic scattering recoil energy production cross section | calculated from neutron energy loss in elastic and inelastic scattering |
-11 | source rate | |
-15 | neutron density | flux multiplied by inverse neutron speed |
-16 | macroscopic total scattering neutron production cross section | |
-30 | temperature majorant cross section | majorant used for rejetion sampling in TMS |
-100 | user-defined response function | see detailed description |
Photon reactions
References
- ^ Herman, M. and Trkov, A. "ENDF-6 Formats Manual." CSEWG Document ENDF-102 / BNL-90365-2009.