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Development a Computer Program for Burn-Up Calculation Using Monte-Carlo Method

Khaki Arani, Mehdi | 2009

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  1. Type of Document: M.Sc. Thesis
  2. Language: Farsi
  3. Document No: 39347 (08)
  4. University: Sharif University of Technology
  5. Department: Energy Engineering
  6. Advisor(s): Ghofrani, Mohammad Bagher; Feghhi, Amir Hossein
  7. Abstract:
  8. Burnup calculations based on the Monte Carlo method have been developed in line with the improvements in computing technology. Nowadays, in the field of nuclear reactor physics, it is possible to perform burnup calculation in a detailed 3D geometry and continuous energy description by the Monte Carlo method. COMB is a fully automated code in DELPHI 7 that links the MCNP4C Monte Carlo transport code with the radioactive decay and burnup calculation code system, ORIGEN2.1. The principal function of COMB is first to transfer one-group cross sections and fluxes from MCNP4C to ORIGEN2.1, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2.1 back to MCNP4C in a repeated, cyclic fashion. The benchmarking process for COMB used four different test cases, representing a variety of burnup scenarios: a PWR pin cell, a PWR fuel assembly, a BWR Fuel assembly and a simple HTR core. The differences between calculated isotopic concentrations in comparison with the results from other codes as well as the experimental data were estimated to be less than 10%. These were primarily due to difference between MCNP cross sections libraries and that of used by other codes; as well as the inherent uncertainty of these libraries. The obtained results are indicative of the fact that this code can safely be used in three-dimensional burnup calculations
  9. Keywords:
  10. Monte Carlo Method ; Monte Carlo N-Particle (MCNP)Code ; Burnup Calculation ; COMB Code ; Origen 201 Code

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