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Showing 1 to 8 of 8 for “"Monte Carlo N-Particle Transport"”.
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Medical room design for a fission converter-based boron neutron capture therapy facility
… calculations, computational methods using the Monte Carlo N-particle transport code, and experimental measurements using the current medical room. The design included the shielded wall of the medical room, a shielding door for entrance into the room, a viewing window, and patient positioning …
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Novel techniques for the detection and localization of attenuated gamma-ray sources
… method uses data simulated with the Los Alamos Monte Carlo N-Particle transport code (MCNPX) for detector efficiency calculations. As a necessary supplement to the process, a program was written to automate all MCNPX simulations by modifying input cards and recording the results for all required …
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Characterization of the Neutron Irradiation System for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories
… of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary …
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Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle
… to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core …
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MCODE-3 : time-dependent depletion isotopics with MCNP-5 and SCALE-6.1
… burnup calculations by combining the Monte Carlo N-Particle transport code, MCNP, with the Oak Ridge Isotope Generation point depletion code, ORIGEN-S. MCNP provides reaction rates and neutron flux in user-specified irradiation material regions. COUPLE, a data-editing code included in …
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Proliferation Resistance of Utilizing Enriched Reprocessed Uranium
… (MARC) model is used. Following the MARC model, Monte Carlo N-Particle Transport (MCNP) computer code was used to simulate fuel burnup of a pressurized water reactor fuel assembly that uses ERU. Both, uranium enrichment and fuel burnup simulations allowed us to estimate the Pu-238 buildup in the …
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Conceptual Development of Remote Monitoring System for Pressurized Water Reactor Spent Fuel Dry Cask Storage Through Neutron and Gamma Transport Simulations
… approach was used to develop the proposed RMS. Monte Carlo N-Particle transport code (MCNP) was employed to develop a dry cask model with 32 SNF assemblies inside. The ORIGEN-ARP, fuel burn-up and depletion code, was used to generate a radiation source-term. A series of MCNP simulations were …
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Comprehensive fuel performance model development and simulation for triso fuel
Item withdrawn by Alexis Thompson (athmpsn1@illinois.edu) on 2011-01-13T19:58:31Z Item was in collections: University of Illinois Theses & Dissertations (ID: 1) No. of bitstreams: 1 Hu_Jianwei.pdf: 5191894 bytes, checksum: 702afe062ec7e3d71c3ef29c7b15f21a (MD5)