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Title: Implementation and Validation of Photon Transport in OpenMC

Technical Report ·
DOI:· OSTI ID:1490825
 [1];  [1]
  1. Argonne National Lab. (ANL), Argonne, IL (United States)

OpenMC is a Monte Carlo particle transport code focused on reactor physics calculations. It stochastically simulates neutrons moving through 3D models using constructive solid geometry and continuous-energy cross sections. Recently, a photon transport capability was added to OpenMC. In reactor simulations, photon transport is needed, for example, to account for photon heating: although the majority of the energy deposited at a fission site comes from the kinetic energy of the fission fragments, a nontrivial portion can be carried away by photons. Therefore, photons produced in neutron reactions must be simulated in order to accurately model energy deposition. OpenMC can now simulate neutron-induced photon production as well as perform pure photon calculations. The four basic photon interactions with matter — coherent (Rayleigh) scattering, incoherent (Compton) scattering, the photoelectric effect, and pair production — are modeled. Secondary processes that can create new photons (i.e., atomic relaxation, electron-positron annihilation, and bremsstrahlung) are also included. The methods implemented here were validated against MCNP6 for a variety of materials and energies.

Research Organization:
Argonne National Laboratory (ANL), Argonne, IL (United States)
Sponsoring Organization:
USDOE Office of Science (SC), Advanced Scientific Computing Research (ASCR)
DOE Contract Number:
Report Number(s):
ANL/MCS-TM-381; 149145
Country of Publication:
United States

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