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**Lecture 6: Source Specification**

Regular lessons: Neutron sources Photon sources Special MCNP techniques for representing neutron and gamma ray sources Equivalent point sources

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HOMEWORK problems

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HW 6.1 Use MCNP to create an equivalent (alpha,n) point source for a 210Po-Be source with the following properties: 1 mg of 210Po evenly distributed in a Diameter=Height=1 cm pellet of Be Plot the spectrum from 0 MeV to 10 MeV in increments of 0.1 MeV

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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 e1 e2 e3 …**

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

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**Equivalent point source**

An “equivalent point source” is found by combining two MCNP runs: The first one is a stand-alone model of the physical source. It should include a Fx1 tally on the outer surface that includes an Ex1 card to collect an energy histogram of the escaping particles of interest The second one includes the source as a point source with: an ERG=Dx on the SDEF card matching SIx H… and SPx cards to reproduce the escaping energy histogram A simple FM xx.xxxx card for the resulting escaping source strength (the sum of the values in the Fx1 tally)

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**Multiplier/cross section response: FM**

Syntax: Description: Provides a constant multiplier to be applied to the tally. Since Monte Carlo is normally done on a per-particle basis, this allows you to include a source strength (or units change). Other use is to put in cross-section dependent response functions to make a tally keep up with particular reaction rates. We will discuss this second option later in the lecture

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HW 6.2 Check your answer in HW 6.1 with a hand calculation using the (alpha,n) properties of the source.

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HW 6.3 Using an “expanded” FM card, find the neutron absorption rate in the Be pellet

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**Multiplier/cross section response: FM**

Syntax: Description: Provides a constant multiplier to be applied to the tally. Since Monte Carlo is normally done on a per-particle basis, this allows you to include a source strength (or units change). Other use is to put in cross-section dependent response functions to make a tally keep up with particular reaction rates.

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**Using MCNP-Provided Response Functions**

The alternate use of the FM card is to use energy dependent values that MCNP knows to get the reaction rates that you want; Cross sections for any reaction in any material covered by the libraries (using ENDF MT numbers) Special “dosimetry” cross sections for special purposes Syntax: FM14:x C mat# reaction# x=particle type C=multiplier (negative means times atom-density of mat#--in which case C is generally the negative cell volume) reaction#=any standard ENDF MT # + any of the special reaction values from Table 3.5 of MCNP manual (See next slide)

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**Multiplier/cross section response: FM**

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HW 6.4 Use MCNP to model a point source of fission gammas from a single fission event. Use the source energy spectrum from Eq. 4.9 in the text. Check it by binning the energies of the photons that escape from an enclosing sphere. Bin from .1 MeV to 10 MeV in increments of 0.1 MeV. Plot it logarithmically and compare (in words) with the figure from the text (reproduced on next slide).

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HW 6.5 Integrate Eq. 4.9 to determine the number of fission gamma rays (per fission event) with energies between 1 MeV and 2 MeV. Check your answer by summing the appropriate bins from HW 6.4.

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