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| Content Provider | The American Society of Mechanical Engineers (ASME) Digital Collection |
|---|---|
| Author | Sun, Kaichao Ames, Michael Newton, Thomas Hu, Lin-Wen |
| Copyright Year | 2013 |
| Abstract | A neutronic analysis of the Massachusetts Institute of Technology Research Reactor (MITR) is performed using state-of-the-art computational tools: the continuous-energy Monte Carlo code MCNP5 and the point-depletion code ORIGEN2.2. These codes are externally coupled by the in-house code package, MCODE (MCNP-ORIGEN Coupled Depletion Program), more recently, it being extended to MCODE-FM (Fuel Management). The latter features automated input file generation, data manipulation, and post-processing of the output data for the fuel cycle analysis, so that it is used to simulate the fuel management of the MITR. MCODE-FM also has an optional criticality search algorithm to simulate control blade movement. The code validation is carried out by comparing the calculated results to experimental data. Two sets of the comparisons are made in the present paper: 1) the Xe-135 reactivity effect during the reactor start-up and shutdown and 2) the thermal and fast neutron flux in an irradiation capsule in the reactor core. Good agreements have been found. The validated MCODE-FM is therefore useful for neutronic analysis and the fuel cycle simulation of the MITR. The time dependent variation of the key parameters, viz. the control blades’ axial position (maintaining criticality) and the fissile inventory in the fuel, is presented. |
| Sponsorship | Nuclear Engineering Division |
| File Format | |
| ISBN | 9780791855829 |
| DOI | 10.1115/ICONE21-16670 |
| Volume Number | Volume 5: Fuel Cycle, Radioactive Waste Management and Decommissioning; Reactor Physics and Transport Theory; Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls; Fusion Engineering |
| Conference Proceedings | 2013 21st International Conference on Nuclear Engineering |
| Language | English |
| Publisher Date | 2013-07-29 |
| Publisher Place | Chengdu, China |
| Access Restriction | Subscribed |
| Subject Keyword | Fuel management Neutron flux Algorithms Simulation Blades Irradiation (radiation exposure) Nuclear fuel cycle Fuels |
| Content Type | Text |
| Resource Type | Article |
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