ENDF reaction MT's and macroscopic reaction numbers
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Serpent uses standard ENDF reaction MTs to identify neutron and photon reactions. The numbers are used with detector response functions, microscopic cross section calculations 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 MTs 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 definitions of some response functions.
Contents
ENDF Reaction MTs
Neutron reactions
MT | Description | Notes |
---|---|---|
1 | total | |
2 | elastic scattering | |
3 | nonelastic | redundant |
4 | total inelastic scattering | redundant (sum over MTs 51 to 91) |
5 | anything | used for lumping together multiple reaction modes |
11 | (n,2nd) | |
16 | (n,2n) | some nuclides miss MT 16 and have only MTs 875-891 instead |
17 | (n,3n) | |
18 | total fission | sum over all fission channels (MTs 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 | redundant |
28 | (n,np) | |
29 | (n,n2α) | |
30 | (n,2n2α) | |
32 | (n,nd) | |
33 | (n,nt) | |
34 | (n,n^{3}He) | |
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 | redundant |
102 | (n,γ) | (102g/102m for transmutation to ground/isomeric state) |
103 | (n,p) | |
104 | (n,d) | |
105 | (n,t) | |
106 | (n,^{3}He) | |
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 | |
202 | total photon production | |
203 | total proton production | |
204 | total deuteron production | |
205 | total triton production | |
206 | total ^{3}He production | |
207 | total α production | |
301 | total heat production | total heating number multiplied by total cross section (note 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 | MTs 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 | MTs 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 | MTs 700-749 can be used to replace MT 105 |
701-748 | (n,t) to excited states | |
749 | (n,t) to continuum | |
750 | (n,^{3}He) to ground state | MTs 750-799 can be used to replace MT 106 |
751-798 | (n,^{3}He) to excited states | |
799 | (n,^{3}He) to continuum | |
800 | (n,α) to ground state | MTs 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 | MTs 875-891 can be used to replace MT 16 |
876-890 | (n,2n) to excited states | |
891 | (n,2n) to continuum | |
1002 | S(α,β) elastic scattering | not an official ENDF MT number |
1004 | S(α,β) inelastic scattering | not an official ENDF MT number |
Photon reactions
Macroscopic reaction numbers
Neutron reactions
Reaction # | Description | Notes |
---|---|---|
-1 | macroscopic total cross section | |
-2 | macroscopic total capture cross section | sum of all reactions that do not produce secondary neutrons |
-3 | macroscopic total elastic scattering cross section | |
-4 | macroscopic total heating cross section | equivalent with the F8 tally in MCNP |
-5 | macroscopic total photon production cross section | |
-6 | macroscopic total fission 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 |
-53 | macroscopic proton production cross section | |
-54 | macroscopic deuteron production cross section | |
-55 | macroscopic triton production cross section | |
-56 | macroscopic He-3 production cross section | |
-57 | macroscopic He-4 production cross section | |
-80 | total energy deposition | combines responses for fission heating, neutron heating based on KERMA coefficients and analog photon heating |
-100 | user-defined response function | followed by a function name corresponding to a function defined using the fun card, response material is omitted |
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 |
-12 | analog photon heating | Energy deposition detector |
-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 | followed by a function name corresponding to a function defined using the fun card, response material is omitted |
-200 | photon dose rate in local material | in Gy/h, using mass attenuation coefficients from NIST data,^{[2]} see detailed description |
-201 | photon dose rate in A-150 Tissue-Equivalent Plastic | Reaction numbers -201 to -248 are reserved for photon dose rates in pre-defined material compositions using same data as -200 |
-202 | photon dose rate in adipose Tissue (ICRU-44) | |
-203 | photon dose rate in air, Dry (Near Sea Level) | |
-204 | photon dose rate in alanine | |
-205 | photon dose rate in B-100 Bone-Equivalent Plastic | |
-206 | photon dose rate in bakelite | |
-207 | photon dose rate in blood, Whole (ICRU-44) | |
-208 | photon dose rate in bone, Cortical (ICRU-44) | |
-209 | photon dose rate in brain, Grey/White Matter (ICRU-44) | |
-210 | photon dose rate in breast Tissue (ICRU-44) | |
-211 | photon dose rate in C-552 Air-equivalent Plastic | |
-212 | photon dose rate in calcium Sulfate | |
-213 | photon dose rate in 15 mmol/l Ceric Ammonium Sulfate Solution | |
-214 | photon dose rate in cesium Iodide | |
-215 | photon dose rate in concrete, Barite (Type BA) | |
-216 | photon dose rate in concrete, Ordinary | |
-217 | photon dose rate in eye Lens (ICRU-44) | |
-218 | photon dose rate in calcium Fluoride | |
-219 | photon dose rate in ferrous Sulfate (Standard Fricke) | |
-220 | photon dose rate in gadolinium Oxysulfide | |
-221 | photon dose rate in gafchromic Sensor | |
-222 | photon dose rate in gallium Arsenide | |
-223 | photon dose rate in glass, Lead | |
-224 | photon dose rate in photographic Emulsion (Kodak Type AA) | |
-225 | photon dose rate in lithium Fluride | |
-226 | photon dose rate in lithium Tetraborate | |
-227 | photon dose rate in lung Tissue (ICRU-44) | |
-228 | photon dose rate in magnesium Tetroborate | |
-229 | photon dose rate in mercuric Iodide | |
-230 | photon dose rate in muscle, Skeletal | |
-231 | photon dose rate in polyethylene Terephthalate (Mylar) | |
-232 | photon dose rate in radiochromic Dye Film (Nylon Base) | |
-233 | photon dose rate in ovary (ICRU-44) | |
-234 | photon dose rate in photographic Emulsion (Standard Nuclear) | |
-235 | photon dose rate in polymethyl Methacrylate | |
-236 | photon dose rate in polyethylene | |
-237 | photon dose rate in polystyrene | |
-238 | photon dose rate in polyvinyl Chloride | |
-239 | photon dose rate in glass, Borosilicate (Pyrex) | |
-240 | photon dose rate in polytetrafluoroethylene (Teflon) | |
-241 | photon dose rate in cadmium Telluride | |
-242 | photon dose rate in tissue-Equivalent Gas (Methane Based) | |
-243 | photon dose rate in tissue-Equivalent Gas (Propane Based) | |
-244 | photon dose rate in testis (ICRU-44) | |
-245 | photon dose rate in tissue, Soft (ICRU Four-Component) | |
-246 | photon dose rate in tissue, Soft (ICRU-44) | |
-247 | photon dose rate in plastic Scintillator (Vinyltoluene) | |
-248 | photon dose rate in water, Liquid |
References
- ^ Herman, M. and Trkov, A. "ENDF-6 Formats Manual." CSEWG Document ENDF-102 / BNL-90365-2009.
- ^ Hubbell, J. H. and Seltzer, S.M. "Tables of X-Ray Mass Attenuation Coefficients and Mass Energy-Absorption Coefficients." (version 1.4). http://www.nist.gov/pml/data/xraycoef/