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FAIL-SAFE SOURCE-DRIVEN FISSION and FUSION-FISSION HYBRID REACTOR CONFIGURATIONS by MONISH SINGH THESIS Submitted in Partial

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FAIL-SAFE SOURCE-DRIVEN FISSION and FUSION-FISSION HYBRID REACTOR CONFIGURATIONS by MONISH SINGH THESIS Submitted in Partial FAIL-SAFE SOURCE-DRIVEN FISSION AND FUSION-FISSION HYBRID REACTOR CONFIGURATIONS BY MONISH SINGH THESIS Submitted in partial fulfillment of the requirements for the degree of Master of Science in Nuclear, Plasma, and Radiological Engineering in the Graduate College of the University of Illinois at Urbana-Champaign, 2013 Urbana, Illinois Master’s Committee: Professor Magdi Ragheb, Director of Research Professor Rizwan Uddin ABSTRACT A source-driven nuclear reactor configuration with a unity infinite medium multiplication factor fission core ( k 1), is investigated for both fission and fusion-fission hybrid systems. Such a configuration is thought to offer a desirable fail-safe reactor alternative in that the loss of the fission or the fusion neutron sources would automatically lead to a shut-down of the system into a stable subcritical state with an effective multiplication factor of less than unity ( keff 1). This is so since the fission core cannot sustain a chain reaction without the presence of the neutron source. A circulating liquid molten salt using the Th-233U fuel cycle, where the fission products are continuously extracted, further contributes to the fail-safe characteristic by avoiding the cooling needed for the decay heat or afterheat after reactor shut-down. Through the extraction of 233 the Pa relatively long-lived (T1 27 days )precursor isotope, and allowing it sufficient time to 2 decay into its 233U daughter, breeding in either thermal or fast neutron spectra is a distinct possibility. The presence of trace amounts of 232U and the strong gamma-emitting 208Tl daughter isotope offers a desirable non-proliferation characteristic for the cycle. As a proof of principle, a simplified analytical one-group neutronics analysis is first attempted for the pure fission core system. This is then supplemented with numerical one-group criticality calculations using an iterative finite-difference methodology. Further, a more detailed continuous energy Monte Carlo neutronics analysis of the fission core reactor driven by a 233U fission neutron source, Deuterium-Tritium (DT) and Deuterium-Deuterium (DD) fusion neutron sources was conducted using the MCNP5 computer code. The first system studied was a spherical reactor core with a unity infinite medium multiplication factor ( ) and surrounded by a reflector. A 232Th and 233U FLiBe molten salt was used as the fuel in the core. The reactor is made critical with the addition of a thin region of 233 FLiBe salt with a spike of fissile material ( U). With a k in the core and total system keff of unity, the flux profile for the system becomes flat, resulting in uniform fuel burnup and power profile. Such a configuration was found to have a conversion ratio of 1.4 in the core. However, 233U production in the core would not be able to replace the 233U consumed in the fissile source ii region without exceeding a 3-5 percent concentration. This may be possibly achieved using other stockpiled fissile materials such as 235U or Pu239 at higher enrichment levels. Alternatively, the fissile source region can be replaced by a fusion neutron source such as from DT or DD fusion. The system studied consisted of a cylindrical core surrounded by a fusion source. It is envisioned that the source could be provided by several cylindrical electrodynamic inertial fusion generators. A small 318 MWth system can be driven by a 22.3 MW DT source or a 9 MW DD source. A DT system would be able to achieve fissile breeding at the expense of requiring an outside source of tritium. Alternatively, a DD system can use a sodium-based molten salt and breed 233U with a doubling time of 9.2 years. The results of the investigation suggest that source-driven systems associated with a molten-salt can be contemplated with substantial fail-safe benefits. Running a subcritical reactor eliminates the need for excessive reactivity control systems and provides safety in a loss of power transient situation. Furthermore, utilizing a fissile neutron source yields beneficial power and flux profiles. Lastly, such systems can breed fissile material and support a future alternative Th-233U thorium fuel cycle. iii ACKNOWLEDGEMENTS I would like to acknowledge and thank my family for their unlimited love, support, and encouragement throughout my undergraduate and graduate studies. I would like to thank Prof. Magdi Ragheb as a thesis adviser, as well as Prof. Rizwan Uddin for their insight and guidance as I developed this thesis work. This work would not have been completed without their time and patience. iv TABLE OF CONTENTS LIST OF FIGURES .......................................................................................................... vii LIST OF TABLES ............................................................................................................ ix CHAPTER 1: INTRODUCTION ............................................................................................................ 1 CHAPTER 2: LITERATURE SURVEY ................................................................................................. 10 CHAPTER 3: FAIL-SAFE REACTOR DESIGN .................................................................................... 28 CHAPTER 4: METHODOLOGY............................................................................................................ 37 CHAPTER 5: FISSILE SOURCE DRIVEN FAIL-SAFE REACTOR ..................................................... 44 CHAPTER 6: DT DRIVEN FAIL-SAFE REACTOR .............................................................................. 50 CHAPTER 7: DD DRIVEN FAIL-SAFE REACTOR ............................................................................. 59 CHAPTER 8: OTHER DESIGN CONSIDERATIONS ........................................................................... 69 CHAPTER 9: DISCUSSION AND CONCLUSIONS .............................................................................. 76 REFERENCES ................................................................................................................. 79 APPENDIX A: ONE-GROUP FINITE-DIFFERENCE CRITICALITY MODEL ...................................... 84 APPENDIX B: MCNP5 CODE INPUT FILES .......................................................................................... 92 v APPENDIX C: TABULATED RESULTS ................................................................................................. 98 vi LIST OF FIGURES 1-1 Energy and material flows in a fusion-fission hybrid ................................................... 5 2-1 Neutron regeneration factor as a function of neutron energy ........................................ 12 2-2 Radiation-dose-rate buildup at 0.5 m from 5 kg spheres of 233U and 239Pu with admixture 232U and higher plutonium isotopes, respectively ........................................................... 16 2-3 Diagram of Polywell operation .................................................................................... 20 2-4 Configuration of the SABR reactor ............................................................................. 21 2-5 Configuration of the LIFE reactor ............................................................................... 22 2-6 Rubbia energy amplifier .............................................................................................. 24 2-7 Accelerator Transmutation of Waste (ATW) Th-U233 fuel cycle .................................. 25 3-1 Geometry of a spherical core with an infinite reflector and a fissile spiked core-reflector interface ..................................................................................................................... 29 3-2 Reactor geometry ........................................................................................................ 33 3-3 Normalized flux vs radius for various criticality conditions (k∞ =1) ............................ 34 3-4 Normalized flux vs radius for various criticality conditions (k∞<1) ............................. 35 3-5 Normalized flux vs radius for various criticality conditions (k∞>1) ............................. 36 4-1 Point detector illustration ............................................................................................ 38 5-1 Flux vs radius for core surrounded by fissile source region.......................................... 45 5-2 Average flux spectrum over incremental radial sections in core region ........................ 46 5-3 Thorium radiative capture cross section vs. energy ...................................................... 47 5-4 Average flux spectrum over incremental radial sections in the core region .................. 48 6-1 Tritium breeding cross sections for 6Li (red) and 7Li (green) ....................................... 53 6-2 Fissile and fusile breeding vs. 6Li concentration .......................................................... 54 6-3 Flux averaged over the core for DT system ................................................................. 55 6-4 Average flux spectrum over incremental radial sections for DT system ............................................................................................... 56 6-5 Radial flux profile for DT system ................................................................................ 57 6-6 Axial flux profile for DT system ................................................................................. 58 7-1 Flux averaged over the core region of the Li and Na molten salt systems .................... 61 vii 7-2 Na absorption cross sections
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