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**Lecture 3:Basic Concepts**

MCNP reinforcement of concepts Introduction to VisEd Advanced source distributions Surface and volume sources Representing particle beams Tally review

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**VisEd Cheat Sheet Start VisEd.**

File->Open (Do not modify input) to choose and open the input file Click “Color” in both windows Zoom in OR Zoom out to get them right As desired: Click “Cell” or “Surf” to see cell numbers Click “Origin” to make the window “sensitive” to subsequent clicks (in either window) Insert origin coordinates to move around

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**VisEd example Inside a box (100x100x100)**

Torus of Rmajor=20, Rminor=5 on floor Cylinder of radius 20, ht 40 on top of torus Sphere of radius 10 centered in cylinder

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**Source Definition: 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

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**Advanced Source Specification**

Source distributions Volumetric sources Surface sources Energy-dependent binning

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**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 Hhistogram =Ldiscrete =A(x,y) pairs interpolated =Sother distribution #’s MCNP5 Manual Page: 3-61

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**Y axis of a distribution: SP**

Syntax: Description: Specification of y axis of pdf for distribution n. option=blankcompletes SI =-ppredefined 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

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**Examples SI2 H 0 5 20 SP2 0 1 2 … SI3 L 1 2 SP3 1 2 SI4 A 0 5 20**

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**Input shortcuts Description: Saving keystrokes MCNP5 Manual Page: 3-4**

Syntax: 2 4R => 1.5 2I 3 => 0.01 2ILOG 10 => 1 1 2M 3M 4M => 1 3J => 1 d d d 5.4 (where d is the default value for that entry)

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**Energy bins: En Syntax: En**

Description: Upper bounds of energy bins (MeV) for tally n MCNP5 Manual Page: 3-90

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**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

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**Particle crossing tally: F1**

Syntax: Description: Tally of current integrated over a surface. Prefixing with ‘*’ changes the units—particles to MeV. Like other tallies, the time dependence is inherited from the source—the code doesn’t care. MCNP5 Manual Page: 3-78

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**Surface flux tally: F2 Syntax:**

Description: Tally of flux averaged over a surface. Prefixing with ‘*’ changes the units—particles/cm2 to MeV/cm2. MCNP5 Manual Page: 3-78

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**Cell flux tally: F4 Syntax:**

Description: Tally of flux averaged over a cell. Prefixing with ‘*’ changes the units—particles/cm2 to MeV/cm2. MCNP5 Manual Page: 3-78

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Point Flux Tally A point flux tally is a special tally that collects the flux at a point Now, of course, neither of the flux tallies that we have studied—collision estimates or track length estimates—could possibly be applied to a POINT This is a very specially designed tally in EVERY source point created and EVERY scattered particle contributes its POTENTIAL for scattering to the point in question

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Point Flux A point flux tally is a special tally that collects the flux at a point The MCNP tally 5 is used to set this up, with syntax: Fx5:n x0 y0 z0 R0 x1 y1 z1 R1 … where: (x0,y0,z0) and (x1,y1,z1) are points where the flux is desired R0 and R1 are the radii around the points where flux contributions will NOT be made An extra bonus that you get from a point flux tally is that the UNCOLLIDED flux (the contribution from particles that come straight from the source) are separately reported

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**Point flux tally: F5 Syntax:**

Description: Tally of flux at a point or ring detector. Prefixing with ‘*’ changes the units—particles/cm2 to MeV/cm2. MCNP5 Manual Page: 3-78

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HW 2.4 Use a hand calculation to calculate both the flux and the current on a 5 cm radius disk lying on the z=0 plane, centered on the origin. For the source use a point isotropic 2 MeV neutron source located at (0,0,10). Assume void material fills an enclosing sphere of radius 30 cm (centered on the origin). Check your calculation with an MCNP calculation (within 1% error)

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HW 2.5 Repeat problem 2.4 with the source located at (0,0,20). Explain why the current/flux ratio is different for the two cases (and why it increases). Check your calculation with an MCNP calculation (within 1% error)

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HW 2.6 Repeat the MCNP calculation of problem 2.4 with the enclosing sphere filled with water, only collecting the uncollided neutrons. Explain why the current/flux ratio is different for the two cases (and why it increases).

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HW 3.1 Find the Legendre coefficients for a 2nd order expansion of e-x, -1<x<1 Create a curve of the approximation vs. the actual curve

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HW 3.2 If you have a parallelepiped volumetric isotropic source with a strength of 100 particles/cc/sec and W=20 cm (x dimension), L=10 cm (y dimension), H=50 cm (z dimension): Find the equivalent surface source if the analyst judges that L is insignificant. Find the equivalent line source if the analyst judges that W is also insignificant; and Find the equivalent point source if the analyst judges that H is insignificant as well. For each of these be sure the source size, placement and strength (in appropriate units) is specified.

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HW 3.3 For each of the four sources in the previous problem (the original cuboid + the three bulleted approximations) create the source in MCNP—with the origin at the center of the original cuboid—and compute the fluxes at the point (250,0,0) using an F5 tally.

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HW 3.4 Use the Appendix H data to give me the appropriate source description for an isotropic 1 microCurie Co-60 point source that is 10 years old. Use a hand calculation to find the flux at a distance of 100 cm Check your calculation with an MCNP calculation (within 1% error) using an F5 tally

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HW 3.5 Use an MCNP calculation of a beam impinging on the small water sample to estimate the total cross section of water for 0.1 MeV, 1 MeV, and 10 MeV photons. Compare your answers to the values in Appendix C of the text.

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