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Validation of DT source term modelling in MCNP and MCUNED codes against SINBAD fusion benchmarks
Kodeli, Ivan Aleksander ; Čufar, Aljaž
Twenty-five of the total of over 100 benchmark experiments included in the Shielding Integral Benchmark Archive and Database (SINBAD) were performed using D-T neutron source, particularly the FNG, ...OKTAVIAN, FNS and IPPE benchmarks. In addition to the characterization of the radiation source, description of the shielding set-up, the instrumentation and the relevant detector measurements, most sets in SINBAD contain also deterministic or Monte Carlo (M/C) radiation transport computer models used for the interpretation of the experiments. Since the existing transport codes, such as the Monte Carlo codes MCNP, SERPENT or TRIPOLI, do not allow explicit modeling of the DT reaction, the MCNP code inputs presently available in the SINBAD database make use of the DT neutron source subroutine. However, this approach is reserved only for users with access to the MCNP source code and furthermore requires frequent updates of the source subroutine each time a new MCNP version becomes available. To avoid these inconveniences, alternative possibilities were tested and validated, such as use of the MCUNED code, an extension of MCNPX, and providing the neutron source energy-angular distribution explicitly using the SDEF cards. Testing of the MCUNED acceleration methods is needed before including them with the present MCNP inputs in SINBAD. The corresponding results are compared and discussed for different Frascati Neutron Generator (FNG) benchmarks.
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