Difference between revisions of "ENDF reaction MT's and macroscopic reaction numbers"

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Revision as of 13:36, 24 September 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.


ENDF Reaction MT's

Neutron reactions

MT Description Notes
1 total
2 elastic scattering
3 nonelastic redundant
4 total inelastic scattering redundant (sum over MT's 51 to 91)
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 redundant
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 redundant
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 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 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 (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,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
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 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 \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 followed by a function name corresponding to a function defined using the fun card

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 equivalent with the *f8 tally in MCNP
-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

  1. ^ Herman, M. and Trkov, A. "ENDF-6 Formats Manual." CSEWG Document ENDF-102 / BNL-90365-2009.
  2. ^ 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/