Speaker
Description
The design of shielding for neutron instruments often requires the reactor model to be decoupled from the instrument model, allowing multiple calculations to be performed without transporting particles from the reactor core during every design iteration. This approach requires an accurate and efficient representation of the neutron and photon fields at the interface between both models.
MCNP provides different approaches for transferring particle distributions between independent calculations. Surface source files (SSR/SSW) preserve complete particle histories but may become impractical when high spatial, energy and angular resolution is required because of the large file sizes involved. Alternatively, source distributions can be reconstructed from surface current tallies and implemented through the SDEF card. However, this approach requires extensive data processing and cannot fully reproduce the angular phase space, since arbitrary azimuthal distributions relative to the local surface normal cannot be represented.
Custom MCNP user routines based on tallyx and sourcex were developed to record and sample discretized surface source distributions directly within the transport code. The source is represented by a set of conditional probability distributions for the spatial, energy and angular variables of the particle phase space. This formulation preserves the correlations between position, energy and direction, including the azimuthal angle.
The proposed methodology is tested against conventional track-based sources and SDEF-generated source distributions. Neutron and photon flux distributions evaluated along the instrument axis in several energy groups are used to compare the results obtained with the three source representations. The comparison is performed for the ASTOR and ANDES neutron instruments of the Argentine Neutron Beam Laboratory (LAHN).
Key Words
LAHN, Radiation transport, MCNP, shielding calculation
| Other | Radiation transport |
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