Editing Input syntax manual (section)

=== set mdep ===

 '''set mdep''' ''UNI VOL N MAT<sub>1</sub> MAT<sub>2</sub> ... MAT<sub>N</sub> ''
          ''ZAI<sub>1</sub>'' ''MT<sub>1</sub>
          ''ZAI<sub>2</sub>'' ''MT<sub>2</sub>
          ...
Sets parameters for calculating homogenized microscopic cross sections. Input values:

{|
| <tt>''UNI''</tt> 
| : universe where universe where homogenized microscopic cross sections are generated
|-
| <tt>''VOL''</tt>
| : volume of the universe [in cm<sup>3</sup>] (3D geometry) or cross-sectional area [in cm<sup>2</sup>] (2D geometry)
|-
| <tt>''N''</tt>
| : number of materials included in the calculation
|-
| <tt>''MAT<sub>1</sub> MAT<sub>2</sub> ... MAT<sub>N</sub>''</tt>
| : material names
|-
| <tt>''ZAI<sub>i</sub>''</tt>
| : nuclide identifier ([[Definitions, units and constants#definitions|ZAI]])
|-
| <tt>''MT<sub>i</sub>''</tt>
| : ENDF reaction [[ENDF reaction MT's and macroscopic reaction numbers|MT]]
|}

<u>Notes:</u>
*The option enables the calculation of homogenized few-group microscopic cross sections for the listed nuclides and reactions.
*The homogenized microscopic cross sections are calculated in the few-group structure for the group constant generation (see [[#set nfg|set nfg]] option).
*The cross sections are always calculated in the actual spectrum of the problem, never with the critical spectrum (see [[#set fum|set fum]] option).
*However, if combined with the implicit leakage correction via MC-Fundamental Mode (see [[#set mcleak|set mcleak]] option), all estimates will be leakage corrected.
*Universe:
**Multiple set mdep cards can be given for a single homogenized universe.
*Materials:
**The listed materials must be enclosed inside the homogenized universe.
*Volumes:
**The calculation requires the material volumes to be correctly defined. For more information, see the detailed description on [[Defining material volumes|Defining material volumes]].
**The parameter <tt>''VOL''</tt> was <tt>''VR''</tt> in versions before 2.1.32 (volume ratio of materials included in micro depletion to total homogenized region).
*Nuclide IDs:
**The nuclide identifiers are entered as ZAI, not ZA. 
**E.g., the ZAI for U-235 is 922350 and the ZAI for Am-242m is 952421.
*ENDF reactions:
**Reaction rates are calculated to all states by default
**Transmutation reactions to ground and isomeric states can be calculated by adding "<tt>g</tt> and "<tt>m</tt>" after the reaction MT.
***E.g., 102m is the capture cross section corresponding to daughter nuclide being in isomeric state.
**Fission reactions corresponding to a specific yield in ENDF can be calculated by adding 1, 2, 3, 4, ... after the reaction MT depending on the fission yield data of the nuclide in the data library. 
***E.g., 181 is total fission cross section corresponding to first fission product yield, (this parameter was different before 2.1.32.
**Sum of all capture reactions can be obtained using MT 101
**Some actinides are missing the total fission channel, and setting the MT to 18 produces sum over MT's 19, 20, 21 and 38 (from version 2.1.29 on).
*Special entries:
***If the number of materials <tt>''N''</tt> is zero, the calculation is carried over all burnable materials.
***If the list of nuclide-reaction pairs (<tt>''ZAI<sub>i</sub>-MT<sub>i</sub>''</tt>) is substituted by "<tt>all</tt>", Serpent will generate a micro-depletion output <tt>[input]_mdep.inc</tt> including all the nuclides and all reactions involved in the calculation at beginning of the  simulation aiming to the extract the constant data (stopping the calculation right after).
*The homogenized microscopic estimates are written in:
** <tt>[input]_mdx[n].m</tt> (for branching: <tt>[input]_mdx[n]b[m].m</tt>) output file, where "<tt>n</tt>" is the burnup step index and "<tt>m</tt>" is the branch index: [[Description of output files#Micro depletion output|micro depletion output]].
** <tt>[input].coe</tt> ouput file, by adding <tt>MDEP_XS</tt> in the [[Input syntax manual#set_coefpara|set coefpara]] list: [[Automated_burnup_sequence#Output|automated burnup sequence/coefficient matrix]] output.
*The microscopic cross section calculation methodology is described in a related article.<ref name="mdep">Rintala, A., Valtavirta, V. and Leppänen, J. ''"Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code."'' Ann. Nucl. Energy [https://doi.org/10.1016/j.anucene.2021.108603 '''164''' (2021) 108603]</ref>
*For practical examples, check the presentation on Microscopic group constants with Serpent<ref name="mdep_ugm">Rintala, A. ''"Microscopic group constants with Serpent."'' 10th International Serpent User Group Meeting, Garching, Germany, October 27-30, 2020 [https://serpent.vtt.fi/serpent/mtg/2020_Munich/Rintala1.pdf UGM 2020]</ref>.