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minor doc improvements
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2 changed files with 5 additions and 5 deletions
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@ -278,7 +278,7 @@ channel. Other absorption reactions like :math:`(n,\gamma)` or
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:math:`(n,\alpha)`, on the contrary, produce no neutrons. There are a few other
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idiosyncrasies in treating fission. In an eigenvalue calculation, secondary
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neutrons from fission are only "banked" for use in the next generation rather
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than being tracked as secondary neutrons from elastic and inelastic scattering
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than being tracked as secondary neutrons as elastic and inelastic scattering
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would be. On top of this, fission is sometimes broken into first-chance fission,
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second-chance fission, etc. The nuclear data file either lists the partial
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fission reactions with secondary energy distributions for each one, or a total
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@ -306,8 +306,8 @@ interpolation law. The number of prompt neutrons released per fission event
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:math:`\nu_p` is also given as a function of incident energy and can be
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specified in a polynomial or tabular format. The number of delayed neutrons
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released per fission event :math:`\nu_d` can only be specified in a tabular
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format. In practice, we only need to determine :math:`nu_t` and
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:math:`nu_d`. Once these have been determined, we can calculated the delayed
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format. In practice, we only need to determine :math:`\nu_t` and
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:math:`\nu_d`. Once these have been determined, we can calculate the delayed
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neutron fraction
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.. math::
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@ -71,12 +71,12 @@ generation, and particle number of the desired particle. For example, to create
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a track file for particle 4 of batch 1 and generation 2::
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settings = openmc.Settings()
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settings.track = (1, 2, 4)
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settings.track = [(1, 2, 4)]
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To specify multiple particles, the length of the iterable should be a multiple
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of three, e.g., if we wanted particles 3 and 4 from batch 1 and generation 2::
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settings.track = (1, 2, 3, 1, 2, 4)
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settings.track = [(1, 2, 3), (1, 2, 4)]
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After running OpenMC, the working directory will contain a file of the form
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"track_(batch #)_(generation #)_(particle #).h5" for each particle tracked.
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