Conference Proceedings Paper
MONTE CARLO CALCULATIONS OF HEAT DEPOSITION IN FUEL AND NON-FUEL MATERIALS IRRADIATED IN THE NATIONAL RESEARCH UNIVERSAL REACTOR (NRU)
33rd Annual CNS Conference - 2012 June 10-13
Sinh Nguyen (Atomic Energy of Canada Limited)
Bruce Wilkin (Atomic Energy of Canada Limited)
The Monte Carlo transport code MCNP can be used for determining the heating power to fission power ratios (HPFPR) and power to coolant ratios (PTCR) for various types of fuel irradiated in NRU, depending on the fuel burnup. The method also applies to heating calculations of materials that contain no fuel. In-house patches to MCNP, QFISS (Fission Q-Values) and DPERT (Direct Cross-Section Perturbation), are incorporated for facilitating the process and dealing better with delayed energy deposition. As a result, this approach provides greater confidence that NRU remains within the license envelope.
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