Monte Carlo burnup analysis code development and application to an incore thermionic space nuclear power system Public Deposited

http://ir.library.oregonstate.edu/concern/graduate_thesis_or_dissertations/rr172173d

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  • Lattice bum-up calculations in thermal reactors are complicated by the necessity for use of transport theory to represent fuel rods, control rods, and burnable absorbers, by many time-dependent variables which must be considered in the analysis, and by geometric complexity which introduces time-dependent, spatial-spectral variations. Representation of lattice structure in a core is further complicated by fuel materials and loading patterns which can be non-symmetric, and by the type of material used as the moderator. The incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI) is an example of such a reactor. In this design, the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional bum-up and fuel depletion analysis codes have been found to be inadequate for these calculations, largely for the reasons mentioned above and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. A more sophisticated codes such as the transport lattice type code WIMS is suitable for the terrestrial commercial reactors. However it lacks some materials, such as ZrH, needed in special applications and it is not capable of performing calculations with highly enriched fuel. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the Thermionic Fuel Elements (TFE) is needed to be developed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron/Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations within the TFEs. This required the modification of the MCNP itself to perform the additional task of accomplishing burn-up calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for burn-up. The of burnable absorber Gadolinium which adds complications both in the physical model and the numerical analysis requires frequent spatial and spectral reevaluations as a function of burn-up. This difficulty is overcome by the application of MCNPBURN by assuming that the burnable poison is uniformly mixed in the fuel. MCNPBURN was benchmarked using a standard Pressurized Water Reactor fuel element against the LEOPARD and WIMS computer codes. The results from MCNPBURN show good agreement with LEOPARD and WIMS. The maximum difference between MCNPBURN and either of the two codes was approximately 9%. The differences can be accounted for by the appropriate treatment of the accumulated fission product. Application of the MCNPBURN for the ATI reactor core, which consists of 165 TFEs and operates at 375 kW of thermal power, showed a system lifetime greater than the projected lifetime of 7 years at full power. The average efficiency is about 5.86% and the change in the overall efficiency over the life time is 0.2%. The percentage of fuel mass burned is estimated to be about 3.6% of the initial mass. Another calculation includes the influence of burnable poisons mixed in the peak pins to flatten the overall core radial power distribution was performed. The efficacy of this change is quite apparent in reducing the power effectively in the peak pins though it may give rise in power elsewhere in the core.
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