Re: Re: [fluka-discuss]: use usrtrack card to obtain reaction counts

From: Francesco Cerutti <Francesco.Cerutti_at_cern.ch>
Date: Thu, 8 Aug 2019 14:03:48 +0200

Hallo,

there are two sets of scoring options in USRBIN (WHAT(1) = 0-8 and 10-18,
respectively) and, as clearly written in the manual, fission scoring
requires the first (fission takes place in a point in space). So, if you
want to score by region, you should use WHAT(1)=2.0 and not 12.0. In
Flair, this appears as 'Region point' rather than 'Region'.

Best regards

Francesco

**************************************************
Francesco Cerutti
CERN-EN/STI
CH-1211 Geneva 23
Switzerland
tel. +41 22 7678962

On Thu, 8 Aug 2019, 李立华 wrote:

>
> Dear Francesco Cerutti:
> I did the simulation as your prompt,but it did not work well.
> the .err file shows:
> *** Activity/fission/neutron balance binnings cannot be track-length!!!
>
> the .inp and .err files can be found in the accessories.
>
> Best regards
>
>> -----原始邮件-----
>> 发件人: "Francesco Cerutti" <Francesco.Cerutti_at_cern.ch>
>> 发送时间: 2019-08-08 15:36:17 (星期四)
>> 收件人: "李立华" <13691416247_at_ciae.ac.cn>
>> 抄送: "FLUKA discussion" <fluka-discuss_at_fluka.org>
>> 主题: Re: [fluka-discuss]: use usrtrack card to obtain reaction counts
>>
>>
>> Hallo,
>>
>> not USRTRACK (which is meant for spectrum scoring), rather USRBIN (per
>> region - to get the fission counts in each region - or
>> Cartesian/cylindrical - to get a spatial distribution of fission density),
>> as indicated in the manual. With the dedicated generalised particles
>> FISSIONS, HE-FISS and LE-FISS.
>>
>> Best regards
>>
>> Francesco Cerutti
>> **************************************************
>> CERN-EN/STI
>> CH-1211 Geneva 23
>> Switzerland
>> tel. +41 22 7678962
>>
>> On Wed, 7 Aug 2019, 李立华 wrote:
>>
>> >
>> > Dear fluka experts:
>> >
>> >     Can the reaction counts such as fission counts be obtained throught using
>> > usrtract card,
>> >
>> > in mcnp the reaction counts can be obtained throught F4 card,for example,the
>> > fission counts
>> >
>> > can be obtained like this:
>> >
>> > F4:n    cell number
>> >
>> > fm4 -v material number -6(total fission cross-section)
>> >
>> > thanks!
>> >
>> >
>> >
>> >
>> >
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>> >
>> >
>> >
>
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Received on Thu Aug 08 2019 - 15:00:41 CEST

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