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Lecture 3:Basic ConceptsLecture 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 SheetVisEd Cheat Sheet1. Start VisEd.

2. File->Open (Do not modify input) to choose and open the input file

3. Click “Color” in both windows

4. Zoom in OR Zoom out to get them right

5. As desired:1. Click “Cell” or “Surf” to see cell numbers

2. Click “Origin” to make the window “sensitive” to subsequent clicks (in either window)

3. Insert origin coordinates to move around

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VisEd exampleVisEd 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 CardSource 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

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Advanced Source SpecificationAdvanced Source Specification Source distributions Volumetric sources Surface sources Energy-dependent binning

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X axis of a distribution: SIX 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 #’sMCNP5 Manual Page: 3-61

1 2 option kn I I ISI

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Y axis of a distribution: SPY axis of a distribution: SP

• Syntax: • Description: Specification of y axis of pdf for distribution

n. option=blankcompletes SI =-ppredefined functionThe 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

1 2 option kn P P PSP

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ExamplesExamples

SI2 H 0 5 20

SP2 0 1 2

…

SI3 L 1 2

SP3 1 2

…

SI4 A 0 5 20

SP4 0 1 2

…

SI5 1 5

SP5 –21 2

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Input shortcutsInput shortcuts

Description: Saving keystrokes MCNP5 Manual Page: 3-4 Syntax:

2 4R => 2 2 2 2 2 1.5 2I 3 => 1.5 2.0 2.5 3.0 0.01 2ILOG 10 => 0.01 0.1 1 10 1 1 2M 3M 4M => 1 1 2 6 24 1 3J 5.4 => 1 d d d 5.4

(where d is the default value for that entry)

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Energy bins: EnEnergy 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 variablesSource 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: F1Particle 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

1 2 3xx : ( )pl S S SF 1

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Surface flux tally: F2Surface 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

1 2 3xx : ( )pl S S SF 2

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Cell flux tally: F4Cell 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

1 2 3xx : ( )pl C C CF 4

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Point Flux TallyPoint 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 FluxPoint 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: F5Point 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

0xx : pl X Y Z RF 5a

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HW 2.4HW 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 2.5HW 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 2.6HW 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 3.1HW 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 3.2HW 3.2If 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 3.3HW 3.3For 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 3.4HW 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 3.5HW 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.