Score only the source neutron undergoing absorption and scattering ,excluding fission neutron.

Thanks for your advice.

I followed your advice, but it didn't work.
If I score BEAMPART 210 Primary (source or beam) particles, the result is
fluence of Primary neutron, without any interaction such as absorption and
scattering.
This is NOT exactly what I want.

Let me clarify my need.

The source neutrons, undergoing absorption, scattering and fission, move to
the detector. I want to calculate the fluence of neutrons in the detector.
But I want to exclude any fission neutrons in my fluence result.

Thanks again.

-----Original Message-----
From: owner-fluka-discuss_at_mi.infn.it
[mailto:owner-fluka-discuss_at_mi.infn.it]
Sent: 2012 17:17
To: sugenghua; fluka-discuss_at_fluka.org
Subject: RE: Score only the primary neutron,excluding fission neutron.

Try to use for scoring in your estimator this general particles:
BEAMPART 210 Primary (source or beam) particles
(See also Fluka Manual page 45)

Cheers

-----Original Message-----
From: owner-fluka-discuss_at_mi.infn.it
[mailto:owner-fluka-discuss_at_mi.infn.it] On Behalf Of sugenghua
Sent: Montag, 20. Februar 2012 02:21
To: fluka-discuss_at_fluka.org
Cc: sugenghua_at_cgnpc.com.cn
Subject: Score only the primary neutron,excluding fission neutron.

Dear FLUKA experts:

Please help me with my problem. Thanks.

In my model, the source particle is neutron, emiting from the fuel
assembly in the middle. Water and iron surround the fuel assembly.
I want to score the fluence of neutron in the iron.

I want to score only the source neutron or called primary neutron,
excluding fission neutron. Namely, I don't want secondary neutron to be
transported and scored. Or, I want to switch off fission. What can I do
with FLUKA?

As I know, in MCNP, there is a NONU card, Fission Turnoff Card.
This card turns off fission in a cell. The fission is then treated as simple
capture.

The following is my thoughts, which I have tried and got no effect.

(1)If WHAT(6) of LOW-NEUT is 0.0,is fission neutron multiplicity
forced to 0?

(2)In user routines MDSTCK(management of the stack of
secondaries),

NUMSEC is number of secondary particles produced in the
interaction, can I just NUMSEC=3D3D0?
Received on Tue Feb 21 2012 - 09:42:38 CET