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An Evaluation of the Net Breeding Capability of Heterogeneously-Fuelled Pressure-Tube Heavy Water Reactors with the Thorium Fuel Cycle

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An Evaluation of the Net Breeding Capability of Heterogeneously-Fuelled Pressure-Tube Heavy Water Reactors with the Thorium Fuel Cycle University of Ontario Institute of Technology An Evaluation of the Net Breeding Capability of Heterogeneously-Fuelled Pressure-Tube Heavy Water Reactors with the Thorium Fuel Cycle By Sourena Golesorkhi A thesis submitted in fulfilment of the requirements for the degree of Doctor of Philosophy in the Faculty of Energy Systems and Nuclear Science Supervised By: Dr. Matthew Kaye Supervisory Committee: Prof. John Froats Dr. Glenn Harvel Dr. Benjamin Rouben December 2015 UNIVERSITY OF ONTARIO INSTITUTE OF TECHNOLOGY Abstract Faculty of Energy Systems and Nuclear Science Doctor of Philosophy An Evaluation of the Net Breeding Capability of Heterogeneously-Fuelled Pressure-Tube Heavy Water Reactors with the Thorium Fuel Cycle by Sourena Golesorkhi The thorium fuel cycle is a promising future option for an alternative to uranium. The pressure tube heavy water reactor (PT-HWR) has for decades been considered capable of achieving net breeding on the thorium fuel cycle, but this capability has not been conclu- sively demonstrated. The goal of this work was to perform reactor physics modelling of the 380-channel, 700 MWe PT-HWR, attempting to achieve a self-sustaining equilibrium thorium cycle with specific focus on operational viability. 233 The DRAGON neutron transport code was used to model ThO2/ UO2 fuel lattice cells. Sensitivity analyses were performed for the fissile nuclide content, specific power, and av- erage fuel temperature. Thorium-based fuels are found to have a strongly negative power coefficient of reactivity, leading to criticality concerns in the event of a prolonged shutdown. Several fuel bundle concepts were modelled in order to determine the most favourable as- sembly for breeding. The results showed no significant benefit to other concepts, therefore the standard 37-element bundle was selected due to its wealth of operational experience and well-known operating margins. Homogeneous fuelling was found to be impractical for breed- ing, and heterogeneous fuel bundles were found to not offer significant improvement while increasing complexity. As a result, heterogeneous core configurations using homogeneous bundles were investigated in full core calculations. The DONJON core physics code was used in conjunction with homogenized cross-sections calculated by DRAGON to model the complete reactor, including reactivity devices and supporting structures. Seven fuelling configurations were simulated and iteratively improved through an empirical process. Ultimately, none of the studied configurations could achieve ii net breeding. The most favourable configuration was found to be a heterogeneous seed and blanket core where the blanket (composed of 1.4 at% 233U) was placed in the central region and the seed (composed of 1.6 at% 233U) was placed in the periphery. Two variants of this configuration (differing on the refuelling scheme used in the blanket) were further investigated with instantaneous power simulations with 10 full power days of refuelling. Both variants were found to abide by the existing license limits on maximum bundle and channel power. The variant using 4-bundle shift in both blanket and seed could tolerate an increase in reactor power to 110% FP while maintaining a comfortable margin to safety. A number of postulated misfuelling events were simulated for both variants, and the responses were found to be controllable by the existing reactivity systems. It is noted that there are significant sources of error in the results of this work. Advances in computational methods as well as better nuclear data for the thorium fuel cycle are required for more accurate predictions. Overall, while the PT-HWR has great potential as a very high converting reactor, based on the results of this work, the existing design cannot operate as a net breeder. From an economic perspective, a gain in power output may be more beneficial than an entirely closed fuel cycle. Acknowledgements I would like to begin by thanking, wholeheartedly, Dr. Matthew Kaye for accepting me when I was a Master of Engineering student, for later encouraging me to transfer into the doctoral program, for allowing me the luxury of independence, and for supporting me the entire time. I have also been privileged with a supervisory committee who are not only incredibly dis- tinguished, but have also been incredibly supportive. I would like to thank my original co-supervisor, Dr. Benjamin Rouben, for his infinite patience and constant mentorship; Professor John Froats, from whom I have been able to learn what it truly means to be an engineer; and Dr. Glenn Harvel, who never neglected to tell me when I was not making sense. I have also been fortunate to benefit from the guidance of several other UOIT pro- fessors whose advice has been critical to my development as a graduate student. I thank Dr. Brian Ikeda for long discussions on everything from hockey to lab safety; Dr. Eleodor Nichita, for entertaining my constant questioning on reactor physics; and Dr. Dan Meneley, for his anecdotes and inspiration. This work would not have possible without the support of the computational reactor physics branch of AECL (now CNL). I would like to thank Blair Bromley for his supervision, mentorship, and friendship; Bronwyn Hyland for inviting me to Deep River; Darren Radford for housing me while I was there; Jeremy Pencer, Geoff Edwards, and Mike McDonald for their help and advice over the years; and Shari McGuirl for putting me in touch with them. I would also like to acknowledge Dr. Guy Marleau, Dr. Alain H´ebert, and the Institut de G´enieNucl´eaireat Ecole´ Polytechnique de Montr´eal,for their development and open dissemination of the DRAGON and DONJON codes. I am indebted to my fellow graduate students in the Faculty of Energy Systems and Nuclear Science: Eugene, Jordan, Margarita, Mo, Nicholas, and William, for ensuring that I did not become a social recluse. I am grateful to Martin, for being a sounding board and helping me navigate the Escherian minefield that is mathematics. I would also like to thank Peter Schwanke, who is now a fellow graduate student but who was the original inspiration for my return to graduate school and ardency for reactor physics. Finally, I owe the completion of this work to the help and encouragement of my family and friends (including those already mentioned). I am infinitely thankful to my better half, Robin, for being my best friend, my caretaker, and my partner in all things. To my mom, my other pillar of support, I will never be able to sufficiently express my gratitude, well at least not in English. iii Dedicated to the memory of my father, Vahid Golesorkhi, who passed away at the inception of this work, but who was with me every step of the way. ارسار ازل را هن تو داني و هن من وين رحف معما هن تو خاني و هن م ن گف هست از پس رپده تگوي من و تو چون رپده ربافتد هن تو ماني و هن من The secrets of eternity are not known to thee nor I And this mystery cannot be read by thee nor I Beyond the veil is speech of I and thee For when the veil is lifted there shall remain neither thee nor I The Rubaiyat of Omar Khayyam Contents Abstract i Acknowledgements iii Contents vi List of Figures ix List of Tables xiii Abbreviations xiv Symbols xvii 1 Introduction 1 1.1 Background .................................... 2 1.1.1 Thorium advantages and challenges .................. 3 1.2 Objectives ..................................... 6 1.3 Original contribution ............................... 6 1.4 Outline ...................................... 7 2 Theory 9 2.1 Neutron transport equation ........................... 9 2.2 Computational approach ............................. 10 2.3 Collision probability method .......................... 10 2.4 Burnup ...................................... 11 2.5 Isotopic depletion ................................. 12 2.6 Neutron diffusion equation ............................ 13 2.7 Fissile nuclide content .............................. 14 2.8 Coolant void reactivity .............................. 15 2.9 Breeding ratios .................................. 15 2.10 Effect of protactinium-233 ............................ 16 vi Contents vii 2.11 Infinite lattice cell ................................ 16 2.12 Time-average and instantaneous power ..................... 16 2.13 Power Ratios ................................... 17 3 Literature Survey 18 3.1 Early work .................................... 18 3.1.1 Canadian experience ........................... 19 3.1.2 International experience ......................... 20 3.2 Modern experience ................................ 22 3.3 Startup case .................................... 25 3.4 Economics ..................................... 26 3.5 Regulatory considerations ............................ 28 3.6 Toolset validation and uncertainty ....................... 32 4 Methodology 33 4.1 Toolset ....................................... 33 4.2 Description of model ............................... 35 4.2.1 Refuelling ................................. 39 4.2.2 Reactivity devices ............................ 39 4.3 Experimental process ............................... 41 5 Lattice cell modelling results 43 5.1 Fresh fuel composition .............................. 44 5.2 Sensitivity to power ............................... 47 5.2.1 Response to power maneuvers ...................... 50 5.2.2 Shutdown transients ........................... 52 5.3 Sensitivity to fuel temperature ......................... 54 5.4 Reactivity coefficients .............................. 56 5.5 Alternative fuel bundle concepts ........................ 58 5.5.1
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