More accurately simulate the neutron flux profile#3653
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More accurately simulate the neutron flux profile#3653OceanNuclear wants to merge 102 commits intomainfrom
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Codecov Report❌ Patch coverage is
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## main #3653 +/- ##
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- Coverage 45.98% 45.65% -0.34%
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Files 123 125 +2
Lines 28550 29198 +648
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+ Hits 13129 13329 +200
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…ng the strength of the macroscopic (self and other) scattering matrix.
…(...)), instead of using complex number to cover both +ve and -ve sign in the sqrt.
…esmos). Next step is to write the program for solving multi-group fluxes.
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…ed classes and functions are now in neutronics_data.
…have to import openmc.
…for ease of maintenace; added tests.
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…nded_boundary, to be self-populating dictionaries (populated by solve_group_n).
…r both flux and current, but not for heating yet.
…its from eV to Joule
…so they're still valid under the new units (J, instead of eV)
…iscontinuity present are all justified (i.e. have physical meaning). This in turn updated the behaviour of some methods, and notably two new methods are added to calculate the heating in a per-layer fashion. Docstring updates as well.
… exist without first re-writing solve_lowest_group
…p while keeping _calculate_mean_energy_and_incident_bin as a staticmethod.
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Snapshot of where I'm at before taking a break for 3 weeks: The current state of this branch is:
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…each NeutronFluxProfile.coefficient[:, n] will contain self.n_groups coefficients rather than n coefficients.
…'s coefficients (i.e. coefficients used to cancel the flux shapes contributed by other groups).
… coefficient length = total number of groups rather than the current group number.
…the programmer won't have to write them out explicitly every time.
…us layer can be written as an affine transformation in linear algebra (y = Ax+b). The method _get_propagation_matrix_and_offset_vec returns A and b. num_coef_pairs is no longer needed so is deleted.
… is recorded as a vector of Kronecka delta(n,g) * flux, where g is the group index matching the incident neutron energy.
…thand to some summations (see _summation_shorthand) for ease of maintenance.
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Description
Calculates the neutron flux profile in the first wall and blanket, using an extremely simplified diffusion-based model of an infinite slab sandwich.
Reason for choosing this model
It's better than not having a neutronics model.
It's also very simplified, so it can be calculated on the fly, without installing any deterministic neutron transport codes as dependency.
Checklist
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