Difference between revisions of "ENDF reaction MT's and macroscopic reaction numbers"
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Revision as of 18:19, 25 February 2016
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,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 | |
102 | (n,γ) | |
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 | Note 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 ^{3}He 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 (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 | 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,^{3}He) to ground state | MT's 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 | 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 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 |
-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 in local material | in Gy, using mass attennuation coefficients from NIST data,^{[2]} see detailed description |
-201 | photon dose in A-150 Tissue-Equivalent Plastic | Reaction numbers -201 to -248 are reserved to photon dose in pre-defined material compositions using same data as -200 |
-202 | photon dose in adipose Tissue (ICRU-44) | |
-203 | photon dose in air, Dry (Near Sea Level) | |
-204 | photon dose in alanine | |
-205 | photon dose in B-100 Bone-Equivalent Plastic | |
-206 | photon dose in bakelite | |
-207 | photon dose in blood, Whole (ICRU-44) | |
-208 | photon dose in bone, Cortical (ICRU-44) | |
-209 | photon dose in brain, Grey/White Matter (ICRU-44) | |
-210 | photon dose in breast Tissue (ICRU-44) | |
-211 | photon dose in C-552 Air-equivalent Plastic | |
-212 | photon dose in calcium Sulfate | |
-213 | photon dose in 15 mmol/l Ceric Ammonium Sulfate Solution | |
-214 | photon dose in cesium Iodide | |
-215 | photon dose in concrete, Barite (Type BA) | |
-216 | photon dose in concrete, Ordinary | |
-217 | photon dose in eye Lens (ICRU-44) | |
-218 | photon dose in calcium Fluoride | |
-219 | photon dose in ferrous Sulfate (Standard Fricke) | |
-220 | photon dose in gadolinium Oxysulfide | |
-221 | photon dose in gafchromic Sensor | |
-222 | photon dose in gallium Arsenide | |
-223 | photon dose in glass, Lead | |
-224 | photon dose in photographic Emulsion (Kodak Type AA) | |
-225 | photon dose in lithium Fluride | |
-226 | photon dose in lithium Tetraborate | |
-227 | photon dose in lung Tissue (ICRU-44) | |
-228 | photon dose in magnesium Tetroborate | |
-229 | photon dose in mercuric Iodide | |
-230 | photon dose in muscle, Skeletal | |
-231 | photon dose in polyethylene Terephthalate (Mylar) | |
-232 | photon dose in radiochromic Dye Film (Nylon Base) | |
-233 | photon dose in ovary (ICRU-44) | |
-234 | photon dose in photographic Emulsion (Standard Nuclear) | |
-235 | photon dose in polymethyl Methacrylate | |
-236 | photon dose in polyethylene | |
-237 | photon dose in polystyrene | |
-238 | photon dose in polyvinyl Chloride | |
-239 | photon dose in glass, Borosilicate (Pyrex) | |
-240 | photon dose in polytetrafluoroethylene (Teflon) | |
-241 | photon dose in cadmium Telluride | |
-242 | photon dose in tissue-Equivalent Gas (Methane Based) | |
-243 | photon dose in tissue-Equivalent Gas (Propane Based) | |
-244 | photon dose in testis (ICRU-44) | |
-245 | photon dose in tissue, Soft (ICRU Four-Component) | |
-246 | photon dose in tissue, Soft (ICRU-44) | |
-247 | photon dose in plastic Scintillator (Vinyltoluene) | |
-248 | photon dose 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/