Skip to content

ENDFReactionTypes.md

Latest commit

 

History

History
179 lines (156 loc) · 12.3 KB

ENDFReactionTypes.md

File metadata and controls

179 lines (156 loc) · 12.3 KB

ENDF/B Reaction Types {#sec:types}

This section lists some of the more useful reactions for use with the FMn input card and with the MCNP`` cross-section plotter. See the relevent section of the MCNP`` manual for more information. The complete ENDF/B list can be found in the ENDF/B manual [MCL95]. The tables and reaction types listed are:

Table Description
1{reference-type="ref" reference="tab:NeutronReactions"} Neutron continuous-energy reactions
2{reference-type="ref" reference="tab:SaB"} $S(\alpha, \beta)$ reactions
3{reference-type="ref" reference="tab:multigroup"} Neutron and photon multigroup reactions
4{reference-type="ref" reference="tab:Photoatomic"} Photoatomic reactions
5{reference-type="ref" reference="tab:Photonuclear"} Photonuclear reactions
6{reference-type="ref" reference="tab:Electrons"} Electron reactions

In each of these tables, the MT column lists the ENDF/B reaction number and the FM column lists special MCNP``6 reaction numbers that can be used with the FM card and cross-section plotter.

The nomenclature between MCNP``6 and ENDF/B is inconsistent in that MCNP``6 often refers to the number of the reaction type as R whereas ENDF/B uses MT, but they are the same. The problem arises because MCNP``6 has an MT input card used for the $S(\alpha, \beta)$ thermal treatment. However, the nomenclature between Monte Carlo transport and deterministic transport techniques can be radically different. The reference [FRA96c] provides more information.

Generally only a subset of reactions is available for a particular nuclide. Some reaction data are eliminated by MCNP``6 in cross-section processing if they are not required by the problem. Examples are photon production in a MODE N problem, or certain reaction cross sections not requested on an FM card. FM numbers should be used when available, rather than MT numbers. If an MT number is requested, the equivalent FM number will be displayed on the legend of cross-section plots.

::: {#tab:NeutronReactions}

MT FM Microscopic Cross-Section Description
1 -1 Total
2 -3 Elastic
16 $(n,2n)$
17 $(n,3n)$
18 Total fission $(n,fx)$ if and only if MT=18 is used to specify fission in the original evaluation.
-6 Total fission cross section. (equal to MT=18 if MT=18 exists; otherwise equal to the sum of MTs 19, 20, 21, and 38.)
19 $(n,f)$
20 $(n,n'f)$
21 $(n,2nf)$
22 $(n,n'\alpha)$
28 $(n,n'p)$
32 $(n,n'd)$
33 $(n,n't)$
38 $(n,3nf)$
51 $(n,n')$ to 1st excited state
52 $(n,n')$ to 2nd excited state
$\cdots$ $\cdots$
90 $(n,n')$ to 40th excited state
91 $(n,n')$ to continuum
101 -2 Absorption: sum of MT=102--117 (neutron disappearance; does not include fission)
102 $(n,\gamma)$
103 $(n,p)$
104 $(n,d)$
105 $(n,t)$
106 $(n,3He)$
107 $(n,\alpha)$
202 -5 Total photon production
203 Total proton production
204 Total deuterium production
205 Total tritium production
206 Total 3He production
207 Total alpha production
301 -4 Average heating numbers (MeV/collision)
-7 Nubar (prompt or total)
-8 Fission Q (in print table 98, but not plots)

: Neutron continuous-energy and discrete reactions. :::

At the time they are loaded, the total and elastic cross sections from the data library are thermally adjusted by MCNP``6 to the temperature of the problem, if that temperature is different from the temperature at which the cross-section set was processed. If different cells have different temperatures, the cross sections are first adjusted to zero degrees and adjusted again to the appropriate cell temperatures during transport. The cross-section plot will never display the transport adjustment. Therefore, for plotting, reactions 1 and -1 are equivalent and reactions 2 and -3 are equivalent. However, for the FM card, reactions -1 and -3 will use the zero-degree data and reactions 1 and 2 will use the transport-adjusted data. For example, if a library evaluated at 300 K is used in a problem with cells at 400 K and 500 K, the cross-section plotter and MT=-1 and MT=-3 options on the FM card will use 0 K data. The MT=1 and MT=2 options on the FM card will use 400 K and 500 K data.

The user looking for total production of p, d, t, , and should be warned that in some evaluations, such processes are represented using reactions with MT (or R) numbers other than the standard ones given in the above list. This is of particular importance with the so-called "pseudolevel" representation of certain reactions which take place in light isotopes. For example, the ENDF/B-V evaluation of carbon includes cross sections for the $(n,n’3\alpha)$ reaction in MT=52 to 58. The user interested in particle production from light isotopes should check for the existence of pseudolevels and thus possible deviations from the above standard reaction list.

::: {#tab:SaB} MT FM Microscopic Cross-Section Description

1 Total cross section 2 Elastic scattering cross section 4 Inelastic scattering cross section

: $S(\alpha, \beta)$ reactions. :::

::: {#tab:multigroup} MT FM Microscopic Cross-Section Description

1    -1  Total cross section

18 -2 Fission cross section -3 Nubar data -4 Fission chi data 101 -5 Absorption cross section -6 Stopping powers -7 Momentum transfers n Edit reaction n 202 Photon production 301 Heating number 318 Fission Q 401 Heating number times total cross section

: Neutron and photon multigroup reactions. :::

::: {#tab:Photoatomic} MT FM Microscopic Cross-Section Description

501 -5 Total 504 -1 Incoherent (Compton + Form Factor) 502 -2 Coherent (Thomson + Form Factor) 522 -3 Photoelectric with fluorescence 516 -4 Pair production 301 -6 Heating number

: Photoatomic reactions. :::

::: {#tab:Photonuclear} MT FM Microscopic Cross-Section Description

     1    Total
     2    Non-elastic
     3    Elastic
     4    Heating
     5    Other
    1005  Neutron production from reaction 5
    2005  Photon production from reaction 5
    9005  Proton production from reaction 5

: Photonuclear reactions. :::

::: {#tab:Electrons} MT FM Microscopic Cross-Section Description

    1   de/dx electron collision stopping power
    2   de/dx electron radiative stopping power
    3   de/dx total electron stopping power
    4   electron range
    5   electron radiation yield
    6   relativistic $\beta^2$
    7   stopping power density correction
    8   ratio of rad/col stopping powers
    9   drange
    10  dyield
    11  rng array values
    12  qav array values
    13  ear array values

: Electon reactions. :::

LANL maintains two electron-transport libraries, EL and EL03. The electron transport algorithms and data in MCNP``6 were adapted from the ITS code [HAL92]. The EL library was developed and released in 1990 in conjunction with the addition of electron transport into MCNP``4; the electron-transport algorithms and data correspond (roughly) to that found in ITS version 1. The EL03 library [ADA00] was developed and released in 2000 in conjunction with upgrades to the electron physics package; these upgrades correspond (roughly) to that of ITS version 3.The MT numbers for use in plotting the cross-section values for these tables should be taken from print table 85 column headings and are not from ENDF.