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.


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 \nu\Sigma_\mathrm{f}
-8 macroscopic total fission energy production cross section \kappa\Sigma_\mathrm{f}
-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

Reaction # Description Notes
-9 majorant cross section Note to developers: check that this really works
-11 source rate Note to developers: check that this really works
-15 photon density flux multiplied by 1/c (Note to developers: check that this really works)
-25 macroscopic total cross section Note to developers: use -1 instead?
-26 macroscopic total heating cross section Note to developers: use -4 instead?
-27 photon pulse-height detector see detailed description
-100 user-defined response function see detailed description (Note to developers: check that this really works)
-200 photon dose local material see detailed description
201 A-150 Tissue-Equivalent Plastic
202 adipose Tissue (ICRU-44)
203 air, Dry (Near Sea Level)
204 alanine
205 B-100 Bone-Equivalent Plastic
206 bakelite
207 blood, Whole (ICRU-44)
208 bone, Cortical (ICRU-44)
209 brain, Grey/White Matter (ICRU-44)
210 breast Tissue (ICRU-44)
211 C-552 Air-equivalent Plastic
212 calcium Sulfate
213 15 mmol/l Ceric Ammonium Sulfate Solution
214 cesium Iodide
215 concrete, Barite (Type BA)
216 concrete, Ordinary
217 eye Lens (ICRU-44)
218 calcium Fluoride
219 ferrous Sulfate (Standard Fricke)
220 gadolinium Oxysulfide
221 gafchromic Sensor
222 gallium Arsenide
223 glass, Lead
224 photographic Emulsion (Kodak Type AA)
225 lithium Fluride
226 lithium Tetraborate
227 lung Tissue (ICRU-44)
228 magnesium Tetroborate
229 mercuric Iodide
230 muscle, Skeletal
231 polyethylene Terephthalate (Mylar)
232 radiochromic Dye Film (Nylon Base)
233 ovary (ICRU-44)
234 photographic Emulsion (Standard Nuclear)
235 polymethyl Methacrylate
236 polyethylene
237 polystyrene
238 polyvinyl Chloride
239 glass, Borosilicate (Pyrex)
240 polytetrafluoroethylene (Teflon)
241 cadmium Telluride
242 tissue-Equivalent Gas (Methane Based)
243 tissue-Equivalent Gas (Propane Based)
244 testis (ICRU-44)
245 tissue, Soft (ICRU Four-Component)
246 tissue, Soft (ICRU-44)
247 plastic Scintillator (Vinyltoluene)
248 water, Liquid

References

  1. ^ Herman, M. and Trkov, A. "ENDF-6 Formats Manual." CSEWG Document ENDF-102 / BNL-90365-2009.