1 Lecture 6: Source Spec ification  Source distributions  Volumetric sources  Surface sources  Energy-dependent binning

7 Source Definition: SDEF Card  SDEF card  For a point source:  PAR=1/2/3 particle type (1/2/3=n/p/e)  ERG=xx Energy of particle (MeV)  POS=x y z Position indicator  Example: 9.5 MeV neutron source at point (1., 4., 5.) SDEF PAR=1 ERG=9.5 POS=1 4 5 SDEF PAR=1 ERG=9.5 POS=1 4 5

8 X axis of a distribution: SI Syntax: Description: The SIn and SPn cards work together to define a pdf to select a variable from. option= blank or H  histogram =L  discrete =A  (x,y) pairs interpolated =S  other distribution #’s MCNP5 Manual Page: 3-61

9 Y axis of a distribution: SP Syntax: Description: Specification of y axis of pdf for distribution n. option=blank  completes SI =-p  predefined function The P values are the y-axis values OR the parameters for the desired function p—and the SI numbers are the lower and upper limits. (Table 3.4) MCNP5 Manual Page: 3-61

10 Source description variables  Commands:  POS=Position of a point of interest  RAD=How to choose radial point  AXS=Direction vector of an axis  EXT=How to choose point along a vector  X,Y,Z=How to choose (x,y,z) dimensions  VEC=Vector of interest  DIR=Direction cosine vs. VEC vector  Combinations:  X,Y,Z: Cartesian (cuboid) shape  POS, RAD: Spherical shape  POS, RAD, AXS, EXT: Cylindrical shape  VEC,DIR: Direction of particle

11 SP card special functions  The corresponding SI card gives the min and max of the variable

12 Examples SI2 H SP … SI3 L 1 2 SP3 1 2 … SI4 A SP … SI5 1 5 SP5 –21 2

13 Cell tracking: CF Syntax: Description: Works with tally type 1, 2, 4, 6, and 7 to separately tally particles that have passed through particular cells of the geometry. MCNP5 Manual Page: 3-99

14 Surface tracking: SF Syntax: Description: Works with tally type 1, 2, 4, 6, and 7 to separately tally particles that have passed through particular surfaces of the geometry. MCNP5 Manual Page: 3-100

15 Input shortcuts  Description: Saving keystrokes  MCNP5 Manual Page: 3-4  Syntax:  2 4R =>  1.5 2I 3 =>  ILOG 10 =>  1 1 2M 3M 4M =>  1 3J 5.4 => 1 d d d 5.4 (where d is the default value for that entry)

16 Time bins: Tn  Syntax: Tn  Description: Create time bins in shakes (10 -8 sec)

17 Energy bins: En  Syntax: En  Description: Upper bounds of energy bins (MeV) for tally n

18 VisEd skill: Plotting source points  Create the input file  Open VisEd to view the geometry  Zoom in or out/translate axis as desired to give you the view you want  Click anywhere inside the view you want to source particles to appear in  Under the main toolbar “Particle display” choose “Plot Particle Tracks”  A window of choices will pop up  Move the new popup window out of the way, if you want  Choose: Number of particles to plot: Play with this to get the density of particles you want Number of particles to plot: Play with this to get the density of particles you want Distance from plot plane: This is how far off the plane the particle can be (into or out of the screen) and still be plotted Distance from plot plane: This is how far off the plane the particle can be (into or out of the screen) and still be plotted Display: Source (For this exercise only. You can make other choices to see collisions, etc.) Display: Source (For this exercise only. You can make other choices to see collisions, etc.)  Click on “Plot_Source” on popup toolbar  Example: EX3A

19 VisEd skill: Plotting cross sections  Create input file  Open VisEd  Click toolbar “Cross Section Plots”  A popup window will appear  Click “Read_Cross_Sections” from popup toolbar  Beside “Nuclide” (you could also plot material cross sections using “Material”):  Enter a number in the box and press “Plot”  It will NOT plot anything, but the error message will tell you which “zaids” it knows about  Pick the one you want and type it into the box (including any suffix, e.g., “.70c”)  From the pull-down menu that is under the “Nuclide” box (it probably says “1 Total (N)” in it) and pick the reaction you want it to plot  Absorption is way down the list at 101  Photon cross sections start at 501  The inelastic neutron cross sections start at 51  Press “Plot” again. It will show up in the active window  Example: EX4A

20 Base case of HW#5  Just using an empty sphere with a source at origin:

21 Base case input deck HW#5 base case c ********************************************************************* c * c Cells * c * c ********************************************************************* imp:n= imp:n=0 c ********************************************************************* c * c Surfaces * c * c ********************************************************************* 1 sph c ********************************************************************* c * c Data cards * c * c ********************************************************************* mode n sdef pos = erg=10 f1:n 1 ctme.25 PRINT

22 HW 5.1 Create the following variations of the base case. Put in the appropriate tallies (or take VisEd screenshots) to check your results. A. Put a PX at to see how many go right B. DIR: Make 75% go to right, 25% to left C. ERG: U235 fission neutron spectrum D. 8 cm cube source centered on (0,0,0) E. 4 cm spherical source around origin

23 Set up a Mesh Tally  This is a mesh of rectangles (you can also do a cylindrical mesh) that the answer will be collected on.  This uses the FMESH card, with the following syntax: FMESHx4:n ORIGIN x0 y0 z0 IMESH x1 IINTS nx JMESH y1 JINTS ny KMESH z1 KINTS nz JMESH y1 JINTS ny KMESH z1 KINTS nz OUT ij OUT ijwhere: (x0,y0,z0) is the lower left corner of the mesh (x1,y1,z1) is the upper right corner of the mesh nx,ny,nz tell how many divisions there are in the mesh in the 3 dimensions OUT ij indicates that you want an (x,y) grid of points (for each z level) (You can use other combinations (e.g., jk would give you a (y,z) grid for each x level) (You can use other combinations (e.g., jk would give you a (y,z) grid for each x level)

24 Set up a Mesh Tally (2)  The mesh itself is put into a file that has..msht added to your file name This is if you use the name=file syntax to run MCNP6. Otherwise look for it among the newer files created. This is if you use the name=file syntax to run MCNP6. Otherwise look for it among the newer files created.  This is how I generally view it in EXCEL: 1. Open EXCEL 2. Open the mesh tally file using an editor that you can cut-and-paste from 3. Go to the line just under the “Tally results” line and copy the entire grid into the EXCEL file 1. Be aware that it may put each row in a SINGLE cell in the column. If so, highlight the column and use “Data” and “Text to columns” and “Finish” to spread out the rows. 2. You also may need to move the ROW of x-dimensions over 1 space to the right 4. Highlight (grab) the entire table from the upper left hand corner to the lower right (including the x and y coordinates) 5. Choose “Insert” “Other charts…” and any of the Surface plots 6. I like to use “Layout” “Axes” “Primary Vertical Axis” to: 1. Change to a logarithmic scale with Base 2 or Base 10 (depending on what looks best) 2. Set the color ranges so the colors represents factors of 2 or I also like to click on “3D Rotation” and play with the view (although I also like the “straight down” view using the colors as 2D isoflux lines

25 Set up a Mesh Tally (3) Test of FMESH tally c cells c surfaces 1 sph rpp c data mode p sdef erg=2 pos = par=2 fmesh4:p origin imesh 500 jmesh 500 kmesh 1 iints 40 jints 40 kints 1 out ij iints 40 jints 40 kints 1 out ij imp:p m

26 Set up a Mesh Tally (4)

27 Set up a Mesh Tally (5)

28 Set up a Mesh Tally (6)

29 HW 5.2  Give me a source-height neutron flux map from a 5 MeV neutron point source in the center of a 10’x10’x10’ room.  There is an identical room (through an intervening wall) next to the room the wall is in. Put in a 3 ft wide void opening in the middle of the wall (floor to ceiling).  Assume that the walls are 6” thick water (1 g/cc).  This is the xy view (the + is the source):