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Maximum Fuel Utilization in Advanced Fast Reactors Without Actinides Separation

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Maximum Fuel Utilization in Advanced Fast Reactors Without Actinides Separation Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation By Florent Heidet A dissertation submitted in partial satisfaction of the requirements for the degree of Doctor of Philosophy in Engineering – Nuclear Engineering in the Graduate Division of the University of California, Berkeley Committee in charge: Professor Ehud Greenspan, Chair Professor Brian Wirth Professor Alice Agogino Fall 2010 Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation Copyright 2010 by Florent Heidet Abstract Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation by Florent Heidet Doctor in Philosophy in Engineering – Nuclear Engineering University of California, Berkeley Professor Ehud Greenspan, Chair The primary objective of this study was to estimate the maximum fuel utilization that is achievable using fast reactors that are designed to operate with fuel reconditioning when the fuel reaches its radiation damage constraint. The primary functions of the fuel reconditioning are to relieve the pressure of the gaseous fission products, replace the clad and reduce radiation induced defects in the fuel. That is, the fuel recycling processes considered cannot be used for the separation of actinides from most of the fission products and to extract plutonium or any other actinide from the fuel. Hence, these recycling processes are highly proliferation resistant and, hopefully, less expensive than processes traditionally considered for used fuel recycling. With the fuel reconditioning recycling, the maximum discharge burnup is dictated by the reactivity of the fuel and is not limited by the material damage constraints. A couple of fuel management strategies were examined: the conventional multi-batch fuel management and a breed-and-burn mode of operation. The two recycling processes examined are an AIROX-like process and the melt-refining process. It is found that using fuel reconditioning it is possible to increase the fuel utilization by up to one order of magnitude with the conventional mode of operation and by two orders of magnitude using the breed and burn (B&B) mode of operation, relative to that achievable in once-trough LWRs. The conventional mode of operation, in which the fast reactor is constantly fed with enriched fuel, enables to achieve a discharge burnup ranging from 52.4% FIMA for a medium size 1200 MWth fast reactor to 65.3% FIMA for a large 3000 MWth fast reactor. With the innovative breed and burn mode of operation only depleted uranium is required for the fuel feed with the exception of the fissile fuel required for establishing the initial criticality. The achievable burnup is up to 57% FIMA in a 3000 MWth B&B reactor and up to 44% FIMA in a 1200 MWth B&B reactor. In order to sustain the breed and burn mode of operation it is necessary to accumulate in the depleted uranium feed an average burnup of at least 20% FIMA, when using metallic uranium fuel alloyed with 10 weight % zirconium in a tight-lattice core. By discharging the fuel at this minimum required burnup and loading it, after reconditioning, into a new reactor along with fresh depleted uranium, it is possible to spawn additional B&B reactors without need for any additional fissile fuel. With this spawning mode of operation the achievable B&B reactor capacity growth rate is 3.85% per year, without need for uranium enrichment capability or actinides separation capability. The energy value of the depleted uranium currently accumulated in the USA, when used in the proposed breed-and-burn - 1 - fast reactors, is equivalent to at least seven centuries of the total 2009 USA supply of electricity, all sources included. Relative to LWR operating with the once-through fuel cycle, the fuel discharged from the B&B fast reactors at ~57% FIMA features, per unit of electricity generated: (a) ~40% the amount of TRU and Pu; (b) ~10% the inventory of 237Np and its precursors; (c) ~12% of the decay heat from TRU; (d) ~28% of the radiotoxicity; (e) ~7% the neutron emission rate; the latter three are measured one year following discharge. The fraction of the fissile isotopes in the discharged plutonium is comparable but the decay heat and neutron emission rate per unit mass of discharged plutonium are nearly half as large. The proposed modes of operation are expected to improve the economics and the proliferation resistance and, hence, may justify sooner deployment of fast reactors. The deployment of the suggested fast reactor system will constitute a significant step forward towards sustainable nuclear energy. However, technologies for fuel reconditioning need be developed and their economic viability need be established. - 2 - Table of Contents List of Figures vi List of Tables xi List of Symbols xiv Chapter 1: Introduction ........................................................................................................... - 1 - 1.1 Traditional fast reactor systems ............................................................................. - 3 - 1.2 Alternative approaches proposed ........................................................................... - 5 - 1.3 The UC Berkeley proposed approach .................................................................... - 6 - 1.4 Objective of this study and organization................................................................ - 7 - Chapter 2: Methodology ......................................................................................................... - 9 - 2.1 Fast reactor Cores Examined ................................................................................. - 9 - 2.1.1 General considerations .................................................................................... - 9 - 2.1.2 Large core ..................................................................................................... - 10 - 2.1.3 S-PRISM size core ........................................................................................ - 12 - 2.2 Fuel recycling processes ...................................................................................... - 15 - 2.2.1 AIROX process ............................................................................................. - 15 - 2.2.2 Melt-refining process .................................................................................... - 15 - 2.2.3 Discussion ..................................................................................................... - 16 - 2.3 Fission gases release with burnup ........................................................................ - 16 - 2.4 Fuel density change with fission products accumulation..................................... - 17 - 2.5 Neutronic Analysis ............................................................................................... - 18 - 2.5.1 MCNP5 1.40 ................................................................................................. - 18 - 2.5.2 ORIGEN 2.2 ................................................................................................. - 20 - 2.5.3 MOCUP and IMOCUP ................................................................................. - 20 - i 2.5.4 Additional modules ....................................................................................... - 21 - 2.6 Thermal hydraulic analysis .................................................................................. - 21 - 2.6.1 Constraints .................................................................................................... - 22 - 2.6.2 Core pressure drop ........................................................................................ - 23 - 2.6.3 Maximum core power determination ............................................................ - 23 - Chapter 3: Preliminary Assessment of the Achievable Burnup in Fast Reactors without Actinide Separation ............................................................................................ - 24 - 3.1 Fast reactor cores examined ................................................................................. - 25 - 3.2 Attainable single batch burnup ............................................................................ - 27 - 3.2.1 Uranium based oxide fuel ............................................................................. - 27 - 3.2.2 Uranium based metallic fuel ......................................................................... - 28 - 3.2.3 Thorium based metallic fuel ......................................................................... - 30 - 3.3 Attainable multi-batch burnup ............................................................................. - 33 - 3.4 Sensitivity analysis ............................................................................................... - 34 - 3.5 Fuel resources implications .................................................................................. - 35 - 3.5.1 Fuel utilization .............................................................................................. - 35 - 3.5.2 Actinides discharged composition ................................................................ - 36 - 3.5.3 Waste characteristics ..................................................................................... - 37 - 3.6 Conclusions and Discussion................................................................................. - 40 - Chapter 4: Achievable Burnup in Fast Reactors Using Conventional Fuel Management ....................................................................................................................
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