Design optimization and analysis of a fluoride salt cooled high temperature test reactor for accelerated fuels and materials testing and nonproliferation and safeguards evaluations

Design Optimization and Analysis of a Fluoride Salt Cooled

High Temperature Test Reactor for Accelerated Fuels and

Materials Testing and Nonproliferation and Safeguards

Evaluations

by Joshua Richard B.S., Nuclear Engineering University of Florida, 2010 S.M., Nuclear Science and Engineering Massachusetts Institute of Technology, 2012

Submitted to the Department of Nuclear Science and Engineering in partial fulfillment of the requirements for the degree of

DOCTOR OF PHILOSOPHY in NUCLEAR SCIENCE AND ENGINEERING at the

MASSACHUSETTS INSTITUTE OF TECHNOLOGY February 2016

© 2016 Massachusetts Institute of Technology. All rights reserved.

Author: ____________________________________________________________ Joshua Richard Department of Nuclear Science and Engineering Certified by: ____________________________________________________________ Benoit Forget (thesis supervisor) Associate Professor of Nuclear Science and Engineering Certified by: ____________________________________________________________ Charles Forsberg (thesis reader) Principal Research Scientist Accepted by: ___________________________________________________________

Chair, Department Committee on Graduate Students

Design Optimization and Analysis of a Fluoride Salt Cooled High Temperature Test Reactor for Accelerated Fuels and Materials Testing and Nonproliferation and Safeguards Evaluations

by Joshua Richard

Submitted to the Department of Nuclear Science and Engineering on September 8, 2015

in partial fulfillment of the requirements for the degree of Doctor of Philosophy in Nuclear Science and Engineering

Abstract

Fluoride Salt Cooled High Temperature Reactors (FHRs) are a new reactor concept that have recently garnered interest because of their potential to serve missions and generate revenue from sources beyond those of traditional base-load light water reactor (LWR) designs. This potential is facilitated by high-temperature, atmospheric-pressure operation enabled by the incorporation of liquid fluoride salt coolants together with solid microparticle TRISO fuel.

Since no FHR has been built, an important technology development step is the design, construction, and operation of a FHR test reactor (FHTR). The FHTR’s strategic goals cannot be satisfied using small-scale experiments or test loops: (1) develop the safety and licensing basis for a commercial plant; (2) demonstrate technological viability and provide operational and maintenance experience; and (3) test alternative fuels, fluoride salt coolants, and structures in an actual reactor configuration. The goals of the FHTR support the development of the commercial FHR, but are different. The programmatic goals for the FHTR drive the specification of the technical design goals: (1) capability to switch between any one of various potential liquid fluoride salt coolants; (2) provide an irradiation facility for accelerated fuels and materials testing.

The first stage of the present work included an exploration and characterization of the available design space for an FHTR. Many different core, reflector, and assembly designs were evaluated to determine configurations that possessed acceptable performance while satisfying all design constraints. This work resulted in a novel prismatic block assembly design termed Fuel Inside Radial Moderator (FIRM), which leverages spatial self-shielding of the fuel microparticles to increase core reactivity by ~10,000 pcm relative to a traditional prismatic block design, enabling operation with any of the proposed liquid fluoride salt coolants. This stage of work served to focus the search space for the application of formal optimization algorithms to further improve the feasible design.

The second stage of the present work involved the development of a methodology to perform full-core optimization of the feasible FHTR design and its implementation into usable software. The OpenFRO (Open source Framework for Reactor Optimization) code implements the Efficient Global Optimization (EGO) surrogate-based optimization framework, which has been successfully applied to aerospace and automotive engineering optimization problems in the past. OpenFRO extends the EGO framework to full-core reactor optimization in the presence of uncertainty, enabling an effective, automated, and efficient approach for early-stage reactor design. OpenFRO’s EGO implementation imposes minimal computational overhead while reducing the number of required high-fidelity simulations for optimization by 96%.

The final stage of the present work involved the identification and analysis of the optimal design of the FHTR. The optimal design was selected based on its capability to provide the best performance across potential salt coolants and power levels. The optimal design achieved irradiation position fluxes 90%-130% greater than the feasible design initially identified, while satisfying all safety and performance constraints.

Acknowledgements

I am so thankful for the many wonderful people who have supported me throughout this work. My advisor, Prof. Ben Forget, was a constant source of encouragement and instruction throughout the research; his advice and insights were exactly what I needed. I’m also thankful for his deft awareness of when to provide direction and when to allow me to find the answers on my own. He struck a great balance between the two, and I’m a better engineer and a better researcher because of it. Thank you, Ben!

I’m also very indebted to my thesis reader, Dr. Charles Forsberg. His vast engineering expertise is matched only by the size of the ideas he has for the future of nuclear power, and his enthusiasm for these ideas makes me excited to pursue the future of the technology. His understanding of the big-picture goals of reactor technology development was instrumental in guiding the development of the FHTR’s goals and purpose. Even more, his stories about the wild things that have been tried in the past and may yet be developed in the future were supremely enjoyable. I will miss hearing them!

Additionally, I’d like to thank Prof. Kord Smith for his helpful and real-world feedback on this thesis; his perspective on the practical considerations of much of the research was invaluable. Also, thank you to Prof. Michael Driscoll for his guidance and support as my Master’s thesis supervisor, and for his service as my thesis defense chair.

To my brother Will: thanks for the laughs and the fun we’ve had. You continually fill my life with joy. You’re the coolest person in the world, and you always will be. I love you. To my mom and dad: I wouldn’t be getting a doctorate if not for your example. Thank you for the love and encouragement over the years! Aunt Liana and Big O, thank you for your wisdom and zest for life. It’s inspired me to pursue the same in Christ. Thanks to Aunt Char for her care, and thank you, Grammy, for your love. I can’t wait to see you again!

A big shout-out to all my friends here at MIT. To the guys in Course 22, thanks for a fun ride. John Hanson (JP Hanny!), John Stempien, Alex Mieloszyk, Nate Andrews, and Nate Gibson: glad we made it out alive! Couldn’t have done it without you. To the guys who wised up and stopped with their master’s degrees, Thomas Roomy, Alex Rehn, Tim Gruber, Tyrell Arment: thanks for the fun in those first few years, and glad to see the cool things you’ve done since. Here’s to many more exciting things in the future!

To my friends and fraternity brothers at UF: go Gators! To Dave Chauncey, a man amongst boys: you are a man after God’s own heart, and I thank Jesus for the passion you’ve shown for Christ’s name and the encouragement it has brought to my life. To the other men of God I’ve been privileged to share life with, Mark Mayleben, Aaron Shoemaker, Jon Cagan, Craig Blocher, TY, Scott Stallings, Erik Yeary, and many more: thank you for the community and brotherhood we’ve shared. May we continue to stir one another up in Christ, contending as one man for the faith of the Gospel!

I’d also like to thank my friends and community at my church here in Boston, City on a Hill church. You’ve been like a family to me these past five years, and such encouragement

years: Joel and Page; Don, Bette, and Myra; James and Laura; Fabrice, Alyssa, Sean, and Jenny; Joe and Kayleen; Sam and Becca; and Eduardo and Lisa. Your service for the Kingdom was life-giving in the midst of much work, and the times we shared were so joyful. I’m also so thankful to everyone in CG’s I’ve come to know: your friendship has made life so much more full! Thanks also to Pastors Bland, Mike, and Fletcher: your care for the church has made all the difference, and it is greatly appreciated!

Thank you, Jesus, for hope, for faith, for the love you’ve poured out into my life. You have the words of life; you are the wellspring of living water in my heart. Thank you for saving me, though I did not deserve it, and still do not deserve all the good you so freely give. Thank you for the experience you have given me here at MIT! You are awesome. I pray that this work, and the work of my life, would be for your glory, to make much of your name. Amen

Table of Contents

Abstract(...(3! Acknowledgements(...(5! Table(of(Contents(...(7! 1(Introduction(...(10! 1.1(Background(and(motivation(...(10! 1.2(Thesis(objective(...(16! 1.2.1!Exploration!and!characterization!of!the!potential!design!space!...!16! 1.2.2!Design!optimization!...!17! 1.2.3!Optimal!design!analysis!...!18! 1.3(Organization(of(the(thesis(...(19! 2(FHR(Technology(Background(...(20! 2.1(Molten(Salt(Reactors(...(20! 2.1.1!The!Aircraft!Nuclear!Propulsion!Program!and!the!Aircraft!Reactor!Experiment!...!20! 2.1.2!The!Molten!Salt!Reactor!Experiment!...!25! 2.2(High(Temperature(Gas(Cooled(Reactors(...(29! 2.2.1!The!Dragon!Reactor!Experiment!...!29! 2.2.2!Peach!Bottom!Unit!1!...!33! 2.2.3!Fort!St.!Vrain!...!40! 2.3(Fluoride(Salt(Cooled(High(Temperature(Reactors(...(45! 2.3.1!Initial!Concept!...!45! 2.3.2!Liquid!Salt!Cooled!Very!High!Temperature!Reactor!(LSUVHTR)!...!45! 2.3.3!ORNL!Advanced!High!Temperature!Reactor!(AHTR)!...!48! 3(Design(Exploration(and(Characterization(...(56! 3.1(Design(goals(and(constraints(...(56! 3.1.1!Performance!targets!...!57! 3.1.2!Feasibility/operability!constraints!...!59! 3.1.3!Safety!considerations!...!61! 3.2(Method(...(63! 3.2.1!Reactor!physics!analysis!...!63! 3.2.2!ThermalUhydraulics!analysis!...!64! 3.2.3!Fuel!temperature!analysis!...!66! 3.3(Development(of(the(preliminary(core(design(...(69! 3.3.1!Review!of!salt!constituent!cross!sections!...!70! 3.3.2!Simulation!results!of!earlyUstage!FHTR!core!using!various!primary!salt!coolants!...!73! 3.3.3!Assembly!redesign:!the!Fuel!Inside!Radial!Moderator!(FIRM)!configuration!...!80! 3.4(Preliminary(core(design(parameters(...(86! 3.5(Preliminary(core(thermal(analysis(...(88!

4(Design(Optimization(...(90! 4.1(Optimization(goals,(design(variables,(constraints,(and(fixed(parameters(...(90! 4.1.1!Objective!function!...!90! 4.1.2!The!design!variables!...!91! 4.1.3!Fixed!core!parameters!...!97! 4.1.4!Design!constraints!...!97! 4.1.5!Goals!and!approach!for!the!optimization!process!...!98! 4.2(EGO(framework(overview(...(98! 4.2.1!EGO!for!nuclear!reactor!design:!the!OpenFRO!Python!package!...!99! 4.2.2!Seeding!the!initial!design!space:!the!Design!of!Experiment!...!100! 4.2.3!Generating!the!highUfidelity!data:!the!case!creation!module!...!107! 4.2.4!Approximating!the!highUfidelity!function:!the!surrogate!construction!module!...!113! 4.2.5!Optimizing!on!the!objective!function!and!searching!for!new!evaluation!locations:!the! optimization!and!search!module!...!121! 4.2.6!Closing!the!iteration!loop:!evaluating!the!convergence!criteria!...!132! 4.3(OpenFRO(performance(evaluations(...(133! 4.3.1!LHS!DoE!construction!and!optimization!performance!...!133! 4.3.2!Surrogate!construction!performance!...!137! 4.3.3!EGO!performance!using!the!global!Differential!Evolution!optimizer!...!140! 4.3.4!EGO!performance!using!the!multistart!LUBFGSUB!optimizer!...!144! 4.3.5!EGO!performance!using!the!multistart!COBYLA!optimizer!...!151! 4.3.6!EGO!performance!with!the!multistart!SLSQP!algorithm!...!155! 4.3.7!Timing!performance!...!155! 5(Design(Analysis(of(Optimized(FHTR(cores(...(159! 5.1(Optimization(design(results(...(159! 5.1.1!Analysis!of!the!GPM!hyperparameters!...!160! 5.1.2!Development!of!a!new!fuel!and!coolant!channel!configuration!in!the!FIRM!assembly:! ‘smallpins’!...!162! 5.1.3!Optimization!using!flibe!and!nafzerf,!core!power!allowed!to!vary!...!168! 5.1.4!Optimization!using!flibe!and!nafzerf,!core!power!fixed!at!20!MWth!...!174! 5.1.5!Optimization!using!flibe!and!nafzerf,!and!flux!normalized!to!power!as!the!objective!function !...!181! 5.2(Core(flux(distributions(...(184! 5.2.1!Radial!flux!distributions!...!185! 5.2.2!Axial!flux!distributions!...!186!

5.4(Shutdown(margin(...(195! 6(Summary(...(198! 6.1(Background(and(motivation(...(198! 6.2(Design(space(exploration(and(preliminary(design(development(...(199! 6.3(Design(optimization(...(201! 6.4(Design(analysis(...(203! 7(Recommendations(for(Future(Work(...(208! 7.1(Optimization(framework((OpenFRO)(development(...(208! 7.2(FHTR/FHR(reactor(design(analysis(...(210! References(...(212! Appendix(A:(Early(design(evolution(of(the(FHTR(...(222! Initial(FHTR(core(design:(22.033(Reactor(Design(Class(...(222! First(Design(Evolution:(Low]Leakage(Small(Square(Cylinder(Core(with(Fixed(Reflector(...(225! Second(Design(Evolution:(Low]Leakage(Small(Square(Cylinder(Core(with(Fully(Replaceable(Reflector (...(227! Third(Design(Evolution:(Low]Leakage(Large(Core(with(Fully(Replaceable(Internal(and(External( Reflectors(...(229! Design(Tweaks:(Updating(Material(compositions(and(increasing(the(number(of(coolant(channels(per( assembly(...(231!

1 Introduction

1.1 Background and motivation

The Fluoride Salt Cooled High Temperature Reactor (FHR) is a new reactor concept that incorporates high temperature coated particle fuel with a high temperature, low pressure liquid fluoride salt coolant. Initially proposed in 2003 by Forsberg, Peterson, and Pickard, FHRs combine various features of previous advanced reactor designs in a new way [1]. FHRs use the high-temperature coated-particle fuel form initially developed for high temperature gas-cooled reactors (HTGRs), but unlike these systems FHRs operate at atmospheric pressure [2]. FHRs use a liquid fluoride salt coolant for heat removal that is essentially the same as the carrier salt of the molten salt reactors operated by Oak Ridge National Laboratory in the 1950’s and 1960’s, except that FHRs have no fuel dissolved in their coolant salt [3]. FHRs have passive decay heat removal during shutdown and accident conditions similar to advanced light water reactor (LWR) and sodium-cooled fast reactor (SFR) systems, and must remain at high temperatures even during shutdown due to high melting point coolants, similar to SFRs [4] [5].

FHRs have gained recent interest as a reactor concept because their high-temperature (700 C) operation enables high thermodynamic efficiency for power conversion, while their atmospheric pressure operation enables cheaper components and precludes depressurizations during accidents, enhancing safety. FHRs have also been proposed to couple to advanced power cycles to serve markets beyond those of traditional baseload electricity generation, which could potentially increase revenue and provide valuable support to future electrical grid operations. Coupling to an open air Brayton power cycle could enable co-firing with natural gas for power peaking operation, which could have significant revenue enhancements [6]. Including process steam production as a bottoming cycle can also improve revenue when electricity prices are low. The potential for incorporating firebrick heat storage to leverage low natural gas prices may enhance revenue even further [7]. The prospect for multi-mission operation of FHRs makes them a potentially attractive option for future nuclear power generation and motivates further research and development into this reactor class.

The US Department of Energy (DOE) recently sponsored a Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) to develop a potential path forward for FHR technology [8]. The Massachusetts Institute of Technology (MIT), University of California at

initiatives. See Figure 1 for a visual representation of the strategy of the IRP for pursuing a path forward for FHR technology.

Figure 1. Visual overview of the DOE NEUP IRP’s strategy for determining a path forward for FHR technology

The top-level goals for the FHR, as envisioned by the IRP, fall into three categories: economics, environment, and safety [9]. The economic goal is to increase revenue by 50%-100% relative to base-load nuclear power plants. The environmental goal is to support the development of a low-carbon electricity grid with emissions-free base load electricity generation combined with economic, efficient, and responsive dispatchable electricity generation. The safety goal is to preclude large-scale fuel failures if a Beyond Design Basis Accident (BDBA) occurs. These top-level goals drive the design of the commercial FHR.

The goals of the commercial FHR seek to translate the top-level goals for FHR technology to a particular reactor design [10]. The commercial FHR goals are: (1) enhanced revenue via a multi-mission design; (2) limited severe accident consequences; and (3) improved nonproliferation and waste characteristics. The commercial FHR satisfies these goals by the design of its power cycle systems, primary coolant and fuel selection, and safety system implementation.

The commercial FHR is a multi-mission system capable of generating revenue from multiple sources, including peak power sales during times of high electricity prices (in a deregulated electricity market), industrial steam sales during times of low electricity prices, and energy storage via heated firebrick during periods of low natural gas prices. These capabilities are enabled by the implementation of a nuclear air-Brayton combined cycle (NACC) on the power conversion side of the commercial FHR plant. The NACC incorporates a natural gas topping

cycle. NACC operates with nuclear heat for base load operation. For peak electricity production the hot compressed air is further heated after nuclear heat addition. This enables producing additional electricity in a load-following mode on top of the nuclear heat’s base load electricity production. As a topping cycle, the incremental heat to electricity efficiency is 66% - greater than a stand-alone natural gas plant. This arrangement also enables effective pairing with intermittent renewable electricity generation that is expected to comprise an increasing share of electricity production on the grid in the future.

A test reactor is an important step in the path forward for FHR technology in the United States. While it is not the only test facility that will support technology development by any means, a test reactor can uniquely provide necessary capabilities for commercialization of FHR systems. However, the goals for a FHTR will necessarily differ from those of a commercial system, as the intended purpose is different; the commercial FHR seeks to compete as a profitable and safe option for energy generation, while the FHTR is intended for development of FHR technology in an actual reactor configuration.

Test reactors may be categorized into one of two classes, depending on their intended mission. Class I test reactors are intended to assist in the development of a new class of reactor technology. The Dragon high temperature gas cooled reactor experiment and the molten salt reactor experiment discussed in Chapter 2 are examples of Class I test reactors. Class II test reactors are general-purpose irradiation machines constructed solely to perform materials and fuels testing in their designated testing locations. Class II test reactors are generally of a type (light water cooled, heavy water or beryllium reflected) already constructed and well understood, and their design is selected such that the testing mission is the primary criterion, with only secondary consideration given to technology development driving the features of the reactor design itself. The Advanced Test Reactor (ATR) at Idaho National Laboratory and the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory are examples of Class II test reactors [11] [12].

Class I test reactors can be further subdivided into two sub-classes. Class I-A test reactors are intended to provide testing flexibility such that various potential configurations can be evaluated for future development. There can be two levels of flexibility, with first-level flexibility inside the core to accommodate multiple types of fuel and coolant for evaluation, and second-level flexibility to accommodate wholesale core support/lattice structural changes, control rod drive swaps, and other significant system-wide modifications. Class I-A reactors can have first-level

Class I-B test reactors are prototypical of a particular reactor design and simulate as closely as possible the characteristics of that design on a smaller scale. Class I-B test reactors place no particular emphasis on flexibility for widely varying fuel, core, and coolant modifications, instead intending to test one or a few similar configurations before moving as rapidly as possible to a full-scale prototypical plant. Class I-B test reactors are often used to perform systems tests, confirm expected operation and safety behavior, and generally provide an accelerated avenue for scaling to larger designs. Examples of Class I-B test reactors include the BORAX series of boiling water reactor experiments directed by Argonne National Laboratory at the National Reactor Testing Station (now Idaho National Laboratory), as well as the Chinese TMSR-SF 5 MWth pebble-bed reactor [14] [15]. See Figure 2 for a visual diagram illustrating the various classifications of test reactors.

Figure 2. Diagram illustrating the classifications of test reactors

A Class I test reactor is a necessary step in the development of any new reactor technology. A Class I test reactor satisfies several technology development missions that cannot be satisfied by test loops or other means [16] [17]:

• Develop the safety and licensing basis for a commercial plant

If the reactor is Class I-A, it can satisfy an additional mission:

• Test alternative fuels, fluoride salt coolants, and structures in an actual reactor configuration

The FHTR, as a technology development pathway, must necessarily be a Class I test reactor. The question, then, is whether it should be Class I-A, prioritizing configurational flexibility, or Class I-B, prioritizing system design that matches a proposed commercial design as closely as possible. The IRP envisions that the FHTR will be a Class I-A test reactor with dual-level flexibility. The motivation for selecting the FHTR to be a Class I-A test reactor stems from the likely ownership of the FHTR, and the requirements that particular owner may have for a test reactor system [18]. History suggests that if the FHTR were to be of American interest, the US government would be at least a partial owner. This is based on the following two considerations:

• Historic government role. All first-of-a-kind test reactors in the United States have been built by the government. The 2005 Energy Policy Act explicitly recognizes that a test reactor is a government responsibility.

• Vendor response. The timeline for developing a new reactor concept, starting at the test reactor stage, is considered too long by US commercial nuclear power vendors [18]. Commercial involvement in construction and ownership becomes possible after the feasibility of the concept has been demonstrated and technology development moves to the pre-commercial design stage.

Potential US government ownership pathways include:

• US Government. The US government, via the Department of Energy, fully funds the design, construction, and operation of the FHTR. This is historically the most common arrangement.

• US Foreign Partnership with another country. The US government can jointly fund the development and construction of an FHTR, sharing the costs and ownership stake with another partner country. One potential partner could be China; the Chinese government is pursuing a large FHR development effort, and the US DOE currently has a Memorandum of Understanding with the Chinese to share information in the research and development of FHR technology. This partnership could be expanded in the construction and operation

Switzerland [19]. More recently, the Jules Horowitz Reactor (a 100 MWth Class II test reactor) is under development by an international consortium including European, Asian, and Middle Eastern countries [20] [21].

• Public-private partnership. The US government and a privately owned reactor vendor could jointly fund the development and deployment of a FHTR. However, the large uncertainties (technical, financial, regulatory, etc.) associated with a first-of-a-kind system make this option more difficult than purely public options. The challenges of managing intellectual property rights are exacerbated by working with entirely new technology. Commercial drivers (electricity sales, radioisotope production) may not be feasibly incorporated into the technology development and testing mission.

Given the lack of private interest in the long timelines to revenue generation from this early stage of development, this option is considered the least reasonable. Vendors are interested in FHR technology, but would be involved later on, as the technology is closer to deployment. The remaining options all share the characteristic that the reactor is majority owned by governments, in particular the US government. US government ownership suggests that the FHTR will be a Class I-A test reactor with duel-level flexibility based on the following considerations:

• Competitive vendors. The US government as a matter of national policy has always sought to support at least two vendors for each advanced reactor concept. Competitive vendors imply alternative designs, requiring a test reactor that can evaluate multiple potential types of key FHR components such as coolants and fuels.

• Government missions. The US government may be interested in FHR nuclear air-Brayton combined cycle (NACC) technology for applications in national defense or in support of current government needs, including remote or grid-independent power generation. The wide range of potential applications for government missions means that testing flexibility will be a valuable component of a test reactor.

• Institutional challenges. The US government has historically pursued different strategies for developing a test reactor. The development of sodium-cooled fast reactors included the construction of the first Experimental Breeder Reactor (EBR-I), followed quickly by the construction of the EBR-II and later the Fast Flux Test Facility (FFTF). However, for pressurized water reactors (PWRs), the Shippingport facility tested three entirely different cores in quick succession, with each core replacing the previous configuration while leveraging the same or slightly modified site infrastructure. Given the recent difficulties the US government has had with funding and managing major projects, the second strategy (which requires fewer steps) is more favorable.

It is important to note here that core performance is necessarily constrained by virtue of the selected class of test reactor. A Class II test reactor solely designed to test materials without regard for development of a particular reactor technology will be able to use the very best reflector and moderator materials in the ideal geometric configuration to maximize irradiation

fluxes, period. However, a reactor designed to operate with a pre-selected technology for the purpose of development will have to abide by the inherent limitations imposed by that selection, even if, as a Class I-A test reactor, it also seeks to perform some testing and evaluation of fuels and other materials in its core. Thus, the overall mission and goal of the test reactor and its resulting classification must be kept in mind when evaluating its relative irradiation performance; comparing Class II materials testing reactors to Class I-A flexibly configurable technology development test reactors misses the point of these systems and is not a reasonable endeavor.

1.2 Thesis objective

This thesis seeks to approach the pre-conceptual design of a Class I-A FHTR in the context of the DOE IRP’s path forward for FHR technology. The design of the pre-conceptual FHTR will be addressed in three stages in this work: (1) a preliminary design will be developed by exploring many different potential core and assembly configurations to characterize the available design space and identify a feasible configuration that satisfies the design goals and associated constraints; (2) a methodology will be developed to provide for rigorous algorithmic optimization of full-core nuclear reactor designs and implemented in a software package; (3) the optimization software will be applied to the preliminary design for several different potential scenarios and the design with optimal performance will be identified and subsequently analyzed in more detail. The following sections provide a brief elaboration of each of these topics.

1.2.1 Exploration and characterization of the potential design space

The first component of this thesis will be to explore potential design options for the FHTR in an effort to understand the relevant performance benefits and tradeoffs. The FHR concept is still young by reactor design standards, having been conceived as recently as 2004. Much of the design space for these systems remains unexplored, particularly for smaller-sized reactors like test reactors. As such, beginning the characterization of the wide-open design space with more manual, exploratory methods in the early stages of design development is the most effective option for understanding general system responses and generating novel design features that would be difficult and costly to obtain via an automated algorithmic design approach.

the feasible design space. In the pursuit of satisfying the constraints imposed by the project goals, new designs will be developed that will benefit both the FHTR and other future FHR designs.

Enabling configurational flexibility of the reactor core and associated systems as a Class I-A test reactor is important for evaluating the potential of different primary coolants, fuel geometries, and structural materials for commercial application throughout the lifetime of the test reactor. This is a two-level requirement. At the facility level, sufficient space is included to enable replacement of the entire reactor core, as was done at the Shippingport reactor . At the reactor core level, the core is designed to allow for testing of alternate fuel forms, fluoride salt coolants, and reflector configurations.

A design philosophy of minimizing technological risk where appropriate (while still meeting all technical goals and requirements) was incorporated throughout the design process. This was done in recognition that high uncertainty still exists around much of FHR technology, so identifying and eliminating some technology risk via the selection of better-characterized features and components will help to obtain a higher probability of project success. High-temperature microparticle fuel has a high level of technology development; it has been in development since the beginning of HTGR design with the Dragon Reactor Experiment, was developed further in the German HTGR fuel program, and is currently undergoing testing and evaluation as part the U.S. Next Generation Nuclear Plant (NGNP) program [22] [23] [24] [25]. Placing the fuel particles in prismatic graphite blocks similar to those employed in General Atomics’ Ft. St. Vrain also leverages previously-operated technology [26]. While these selections may be reconfigured in the future, the initial core will benefit from limiting the uncertainty associated with using untested advanced fuel materials and construction.

This thesis objective will obtain a preliminary FHTR design that meets the specified design goals within all relevant constraints. This preliminary design focuses the search space for the optimization process, which helps to obtain convergence for the rigorous optimization framework discussed in the next objective.

1.2.2 Design optimization

Once the design space has been initially explored and a feasible design has been obtained, it is appropriate to then apply a formal design optimization framework to improve the design by simultaneously varying important design features in search of a configuration that maximizes (or minimizes) some objective function of interest. This algorithmic approach to design optimization is particularly relevant for design variables that are continuous (or approximately continuous) and which possess particularly strong interdependency effects. By automating the tedium of varying these continuous variables and tracking the sensitivity of the response of the

desired objective function, as well as capturing the effect of varying multiple design variables simultaneously, formalized optimization methods can obtain an improved design configuration with less time and effort required by the designer. Additionally, many of these optimization methods can explicitly incorporate design bounds and constraints into their process, such that they are accounted for as the process proceeds, rather than checked after-the-fact once an improved design is obtained.

However, there are complications in introducing formal optimization methods into reactor design problems. One key issue is that, depending on the size of the design area to be searched, a design optimization that incorporates many design variables can require many simulations to obtain an improved design solution. If these simulations are time-intensive (as modern Monte Carlo neutron transport calculations often are), then the optimization framework may take an unacceptably long time to obtain a solution. Employing a surrogate-based approach to approximate the response of the high-fidelity simulations during the optimization process can be an effective method of reducing the time-intensity of this process while still obtaining an effectively improved design.

Surrogate-based optimization (SBO) methods apply data interpolation and regression methods to high-dimensional nonlinear data sets to approximate the response of the high-fidelity objective function (such as core reactivity or irradiation flux) between locations of actual high-fidelity information. Thus, instead of having to re-run the high-fidelity simulation code for each evaluation location of the optimization algorithm, only a small subset of points are evaluated using the high-fidelity tool and the values in the ‘gaps’ between data points are estimated using statistical techniques. Since the optimization algorithm can require thousands of evaluations to identify function extrema but an effective surrogate can be constructed with tens or hundreds of data points, many potential expensive calculations can be saved when employing a surrogate-based approach.

Work to satisfy this objective will include the development of a methodology and implementing software that adapts a suitable surrogate-based optimization framework for pre-conceptual stage reactor design. Performance aspects such as the convergence of the framework, importance of including statistical uncertainties in the prediction, and selection between various possible optimizers will be evaluated. The software will then be applied to the preliminary FHTR design for the final thesis objective.

optimized design, and the associated improvement in the objective function value obtained by applying the optimization framework to the baseline design. Given that the FHTR is intended as a Class I-A flexible-operation technology development test reactor, several different design configurations will be optimized and compared to each other to characterize their relative ability to switch between different salt coolants and power levels. Also useful is evaluating the importance of the various constraints applied to the problem, determining which constraints are active (i.e. actively impacting the final design variable vector) and which are inactive (not impacting the final vector).

Other interesting aspects of the FHTR’s performance will be analyzed. This includes the flux and power distributions (both radially and axially) present throughout the core, the irradiation fluxes for the large and small irradiation positions, and the associated flux spectra. The shutdown margin associated with all control rods inserted will be evaluated to verify that safe shutdown can be achieved. These additional analyses will help to provide a fuller picture of the reactor’s performance.

Analyzing the optimized FHTR’s proliferation implications relative to the baseline design is also valuable, as the optimization should result in a reduced amount of attractive material loaded into the core. Characterizing safeguards-relevant signals from the spent nuclear fuel of the FHTR will help to understand how measurements will compare to those of existing reactors, and whether existing methods are sufficient for characterizing the spent fuel composition accurately. Since FHRs (and particularly the small FHTR) are a relatively new reactor concept, these analyses will be valuable in helping to better understand how to reliably safeguard the nuclear materials utilized in these systems.

1.3 Organization of the thesis

The technical content of the thesis is organized around the three objectives identified in the previous sections. A background chapter (Chapter 2) first serves to describe the history of reactor technology development that led to the development of the FHR and influences the design choices made during the technical analysis. The first results chapter (Chapter 3) seeks to satisfy the first thesis objective: to characterize the available design space of FHTR systems by evaluating many different large-scale core, assembly, and reflector configurations in pursuit of an acceptable preliminary design. The second results chapter (Chapter 0) presents the methodology and implementing software developed for early-stage full-core generic nuclear reactor optimization, including a discussion of the theory behind the program and performance results for various optimization options made available in the code. The final results chapter (Chapter 0) presents the application of the optimization tool to the preliminary FHTR design to obtain an optimized design configuration, with further analysis of the optimized design’s performance and nonproliferation characteristics. The thesis concludes with a summary (Chapter 0) and recommendations for future work (Chapter 0).

2 FHR Technology Background

The fluoride salt cooled high-temperature reactor (FHR) concept incorporates key design features of previously investigated reactor designs. The liquid fluoride salt coolant employed by the FHR is chemically similar to the fluoride carrier salts of the liquid-fueled Molten Salt Reactor designs investigated by Oak Ridge National Laboratory (ORNL) beginning in 1949. The solid coated-particle fuel embedded in a graphite matrix utilized by the FHR was originally developed for gas-cooled very-high-temperature reactor (VHTR) designs. This chapter includes a discussion of these pre-existing reactor designs and how their development impacts the FHR design.

2.1 Molten Salt Reactors

Molten salt reactors (MSRs) are nuclear fission power systems in which the actinide fuel is dissolved in a liquid salt coolant flowing through the primary system [3]. Typically designed to operate with a thermal neutron spectrum, the fuel passes through moderating material in the designated ‘core region’, thereby localizing fission to this section of the flow loop. MSRs are thus liquid-fueled reactors, where the fuel is not solid or fixed in-place in the core, which stands in contrast to most reactor designs, which use solid fuel forms. A brief history of MSR development follows, as MSR technology development forms part of the basis for the FHR concept.

2.1.1 The Aircraft Nuclear Propulsion Program and the Aircraft Reactor Experiment

The MSR-type design was first proposed during the Manned Aircraft Nuclear Propulsion (ANP) program as a means of powering a jet engine for a nuclear bomber [27]. The U.S. Air Force desired a propulsion system that could operate for a significant period of time and travel long ranges before refueling, and given the concurrent effort to develop a nuclear-powered submarine, decided to investigate the possibility of a nuclear reactor-powered aircraft propulsion system. An image of a reactor test for such a system is shown in Figure 3.

Figure 3. Photograph of the HTRE-3 nuclear-powered jet engine test at the National Reactor Testing Station (now Idaho National Laboratory) [28]

The AEC contracted the Oak Ridge National Laboratory in 1949 to begin researching potential reactor designs. The principal design goals of an aircraft reactor were identified as maximizing operating temperatures and maximizing the power density of the reactor [29]. This first point is also particularly applicable to FHR systems, as these systems have also been proposed to couple to air-Brayton-cycle turbines. This consideration led to the development at ORNL of high-temperature primary coolants composed of liquid fluoride salts. Liquid fluoride salts were selected as the carrier liquid coolant material in the ANP project because they could operate at high temperatures while remaining at low pressure, could withstand operation in a high-radiation field, and could chemically support dissolving the fuel material into the coolant [30]. While the FHR need not consider the fuel’s solubility in the coolant (being a solid-fuel reactor), it is indeed relevant that the coolant may operate in a radiation field at high temperatures and low pressures, as this characteristic enables coupling with a high-temperature air-Brayton cycle power conversion system while remaining at low pressure in the primary loop, which is a significant cost and safety benefit. Thus, many of the reasons that made liquid fluoride salts desirable for selection in the ANP program also make them desirable for use in FHR reactors.

Among many contributions of the ANP program to salt-cooled reactor development was the design, construction, and brief operation of a small test reactor, the 2 MWth Aircraft Reactor Experiment (ARE) [31]. The ARE was initially intended to be a solid UO2-fueled thermal

reactor, moderated by BeO and cooled by liquid sodium. Unfortunately, this design was found to have a positive temperature coefficient of reactivity, and so the solid fuel idea was dismissed.

However, the BeO moderator blocks had already been ordered (see Figure 4), so within those dimensions the idea for a liquid-fuel reactor was explored.

Figure 4. BeO moderator blocks of the Aircraft Reactor Experiment at ORNL [31]

The initial liquid-fuel design proposed to use fluoride-salt fuel in stagnant (non-flowing) sealed tubes in the core, with sodium coolant flow through the moderator blocks to remove heat. However, a characterization of the in-core temperature distribution resulting from this arrangement revealed that very high thermal gradients existed in the stagnant fluoride fuel material, such that the peak fuel temperature was “dangerously near the boiling point of the fuel”, due to the very low thermal conductivity of the fuel salt [31]. Thus, the reactor was changed to a circulating-liquid-fuel design.

The fluoride carrier salt selected for this reactor was NaF-ZrF4 (59.5 mol% NaF, 40.5 mol%

ZrF4, also called nafzerf). Upon initial startup, molten 2NaF-UF4 was added to the clean carrier

salt until the nominal fuel salt composition of 52.8 mol% NaF-41 mol% ZrF4-5.7 mol% UF4 was

Figure 5. Layout of Inconel tubes positioned in the BeO moderating blocks that comprised the ARE core [31]

Being entirely contained in the metallic tubing, the molten salt fuel never directly contacted the BeO moderator blocks, protecting them from chemical dissolution in the liquid fluoride salt melt. In addition to the multi-pass molten-salt fuel flow through the core, molten sodium was flowed through the core’s moderator block structure in a once-through mode to reduce temperature gradients in the core vessel, provide additional heat removal from the BeO moderator core structure, and fill interstitial gaps with a non-fuel fluid at a higher pressure than the fuel fluid to both prevent fuel leaks from remaining in the core structure and to facilitate improved heat transfer in the moderator structure [29]. Space coolers were also placed in the containment well to cool the air volume surrounding the core, which were necessary for removing the additional waste heat transferred to the air by the resistance heaters that surrounded the core vessel and associated coolant piping [31].

Nafzerf was selected as the carrier salt for the ARE because “the materials were readily available, nontoxic, and not too expensive” [32]. The designers recognized that while nafzerf did have a higher melting point (504 °C) than 2LiF-BeF2 (aka flibe, melting point of 459 °C) and a

higher parasitic neutron absorption rate, the aforementioned advantages of nafzerf made it preferable at the time. An additional complicating factor in utilizing flibe was the difficulty (in 1954, when the ARE was being constructed and operated) in obtaining lithium enriched to very high fractions of Li-7 (99.99 at% and above), which is necessary for achieving an acceptable

neutron economy and low tritium generation rate since Li-6 has a large (~3000 barns) thermal neutron (n,t) absorption cross section [33].

Some practical problems with nafzerf were encountered, however. One significant operational difficulty was that the ZrF4 component of the binary salt was relatively volatile, such that it

would sublimate into the cover gas system of the reactor in large quantities. The issue was such that “ZrF4 snow” (so named due to its appearance) traps were designed to capture this material,

adding somewhat to system cost and operational complexity [34]. These traps sought to either condense ZrF4 vapor as it traveled through the cover gas system or cause it to chemically react

with a bed of Al2O3 pellets, as shown in Figure 6. Such mitigation measures did prove effective,

and did not preclude successful operation.

Figure 6. ZrF4 ‘snow’ traps used designed as part of the ARE to capture ZrF4 sublimated

particles in the reactor cover gas system [34]

Another limitation of the ARE was the use of the aforementioned BeO blocks as the primary moderator material. Although beryllium (whether utilized in oxide BeO or metallic Be form) makes an excellent moderator and reflector material, it is very brittle, such that even relatively

Despite these and many other challenges, the ARE was successfully constructed in 3 years (1951-1954) and operated for 96 MW-hr at the end of 1954 [29] [35]. As the first circulating-molten-salt fueled reactor, it provided invaluable design, construction, and operational experience that could not have been obtained by calculations alone. The ARE test was successful enough that, after the end of the ANP program, engineers at ONRL would apply their experience with the ARE to a reactor design focused on civilian nuclear power production: the Molten Salt Reactor Experiment (MSRE).

2.1.2 The Molten Salt Reactor Experiment

Throughout the development of molten-salt fueled systems in the Aircraft Nuclear Propulsion program, the ORNL researchers recognized the potential of such systems for commercial electricity production [36]. The first substantial effort to investigate the potential configurations of a molten salt reactor for commercial civilian applications was started in 1956 by H.G. MacPherson. This investigation’s exploration of several design concepts would influence the design decisions made during development of the next molten-salt-fueled test reactor constructed at ORNL, the MSRE.

Several broad types of molten-salt-fueled reactor were considered initially. The concepts could be categorized as either single-region (single-fluid) reactors, which have a single primary molten salt loop containing the fissile fuel salt, or two-region (two-fluid) reactors, where the driver salt contains the fissile fuel salt and flows through the designated core region, where it is surrounded by a fertile fuel salt in a separate flow loop that absorbs the leakage neutrons and breeds additional fissile material [37]. MSRs were considered suitable for thermal-spectrum breeding because they can dissolve meaningful amounts of thorium in the fluid, unlike aqueous homogeneous reactors or liquid metal slurry reactors [3]. Thorium solubility was important for obtaining thermal-spectrum breeding in a liquid-fueled reactor because the thorium-232/uranium-233 fuel cycle has better breeding performance in a thermal spectrum than the uranium-238/plutonium-239 fuel cycle [36].

The two-fluid model was expected to achieve better breeding performance relative to the single-fluid model [38]. The conversion ratio of the two-single-fluid core was estimated as 1.05, while the conversion ratio of the single-fluid core was estimated as 1.0, making it a breakeven system. However, to reduce the technical risk of the first-of-a-kind system, early ORNL commercial molten-salt reactor designs opted to use a single-fluid configuration, and later incorporate the experience with single-fluid systems to enable the development of two-fluid reactor systems with better breeding performance [37].

Another design consideration was the choice to have the single-fluid reactor’s core take either a homogeneous liquid configuration or a heterogeneous configuration with the molten fuel flowing through an unclad graphite moderating structure. The heterogeneous graphite configuration was

preferred for breeding, as the solid graphite structure provided a greater atom density of moderator than the fluoride salts. However, the earliest designs used a homogeneous core configuration because the compatibility of the graphite in the molten salt was not yet demonstrated [38].

The decision to focus on maximizing the fuel utilization by increasing the conversion ratio or striving to achieve a breeding gain, together with the initial decision to use a homogeneous core, directly impacted the decision to use 2LiF-BeF2 (flibe) as the carrier salt for the UF4 fuel salt.

With a homogeneous core, it was important to use light materials as the salt constituents in order to increase the moderating power of the salt to achieve a more thermal spectrum. Likewise, reducing the parasitic neutron absorption in the core as much as possible was important in meeting this goal, and thus the salt constituents needed to have both good moderating powers and moderating ratios. Flibe has a moderating ratio of 60, while NaF-ZrF4 (nafzerf) has a

moderating ratio of only 10 [39]. In a thermal spectrum, nafzerf also has approximately 3x the parasitic absorption of flibe. Thus, molten-salt fueled reactors designed after the ARE all used flibe as the carrier salt for the fuel, given that these designs sought to maximize the fuel utilization and breeding potential of the reactor system.

After exploring several potential design concepts, it became clear that these designs differed enough from the ARE design such that a new molten-salt fueled test reactor was needed. By this time, experiments with graphite showed that it could successfully operate in the salt environment, as the salt did not wet the graphite and did not penetrate into its pores provided the pore size was small [36]. Thus, the design of the Molten Salt Reactor Experiment (MSRE) was of a small, single-fluid, graphite-moderated system operating at a power level of 8 MWth. Design of the MSRE began in 1960 and construction began in 1962 [36]. The MSRE reached first criticality in 1965 and operated at full power beginning in December 1965. A diagram of the MSRE core vessel can be seen in Figure 7 and a close-up of the core lattice design can be seen in Figure 8. Nominal MSRE operating parameters are provided in Table 1.

Table 1. Design data for the MSRE [40] [41]

Power 8 MWth

Fuel material LiF-BeF2-ZrF4-UF4 (65 mol%-_{29.1 mol%-5 mol%-0.9 mol%)}

Fuel U-235 enrichment (first operational phase) 33%

Core inlet temperature 635 C

Core outlet temperature 663 C

Operating pressure 138 kPa

Design pressure 345 kPa

Active core height 1.63 m

Graphite core diameter 1.40 m

Total core volume 2.55 m3

Core graphite volume 1.98 m3

Core fuel volume 0.566 m3

Core average power density 14 kW/L

Core peak power density 31 kW/L

Number of fuel/coolant channels 1140

Fuel channel size (length x width) 3.05 cm x 1.02 cm

Nominal volumetric core flow rate 0.0757 m3/s

The MSRE’s nominal operating configuration was with LiF-BeF2-ZrF4-UF4 (65 mol%-29.1

mol%-5 mol%-0.9 mol%) fuel salt using uranium enriched to 33% U-235 [40] [41]. The core fuel temperature rise was 28 °C at an average power density of 14 kW/L, with a peak power density of 31 kW/L, indicating high radial and axial peaking. The estimated peak neutron flux (which occurs in the graphite moderator matrix structure) was estimated as 3.3E+13 n/s/cm3, and the estimated fuel temperature coefficient of reactivity was -5.90 pcm/K [42]. The measured fuel temperature coefficient of reactivity was -8.82 ± 4.14 pcm/K [43].

This initial system configuration operated without issue until 1968, which marked the end of the first phase of operation after 9000 equivalent full-power hours [41]. The first phase demonstrated that the MSRE was able to operate at high temperatures (920 K and above) reliably (critical 80% of the time) without excessive corrosion of structural materials, that xenon could be successfully removed from the salt during operation, and that activated equipment could be safely and effectively replaced without harmfully exposing maintenance workers.

The second phase of MSRE operation began with the addition of a chemical processing capability to the primary salt system, such that the U-235 in the fuel salt could be removed and replaced with U-233. On October 2, 1968, the MSRE was taken to critical on U-233 fuel, the first reactor to do so. The MSRE successfully operated on U-233 for 2500 equivalent full-power hours until December 1969, at which point the fuel salt was drained into holding tanks and the reactor was mothballed [44].

2.2 High Temperature Gas Cooled Reactors

Coated particle fuel was first developed for use in high-temperature gas reactors (HTGRs). These reactor systems are designed for operation with core coolant outlet temperatures from 750-1000 °C, which required the development of a novel fuel form capable of operating at higher temperatures than the traditional metal-clad uranium oxide fuel incorporated by light water reactors and early gas-cooled reactors. As this high temperature coated-particle fuel is also utilized by fluoride salt cooled high temperature reactors, it is illustrative to trace a portion of its technological development and its impact on the core design of the high temperature gas reactors that previously incorporated it.

2.2.1 The Dragon Reactor Experiment

Given the experience in the United Kingdom with the design and operation of the Magnox and AGR gas-cooled reactors, the UK Atomic Energy Authority was very interested in developing gas-cooled reactor systems capable of even higher temperature, higher thermal efficiency operation [22]. Joined by a consortium of European countries under the auspices of the Organization for European Economic Cooperation (now called the Organization for Economic Cooperation and Development, OECD), the UK AEA established the Dragon Project to perform research and experiments in support of developing high-temperature, helium-cooled, microparticle-fueled, graphite-moderated reactor systems, referred to as High Temperature Gas Reactors (HTGRs) [19]. A test reactor, the Dragon Reactor Experiment, was a key component of the research and development program.

The Magnox and AGR gas-cooled graphite-moderated reactor systems developed in the UK used CO2 as a coolant and had relatively conventional fuel forms [45]. The Magnox reactor used

uranium metal fuel clad in a magnesium alloy known as Magnox (hence the reactor’s name). The AGR used UO2 fuel clad in stainless steel. Magnox reactors operated with a core outlet coolant

temperature of 412 °C, while the newer AGRs operated with a core outlet coolant temperature of 675 °C. The next phase of gas-reactor technology development, the HTGR systems, sought to operate at core outlet coolant temperatures above 750 °C, necessitating the development of fuel and assembly configurations that eliminated metallic components in the high-temperature active core region. The devised solution was the creation of an entirely new fuel form: microparticle fuel.

Rather than have fuel take the form of metallic uranium rods or ceramic pellets inside of metallic cladding, microparticle fuel instead incorporates thousands of sub-millimeter diameter fuel kernels coated in one or more protective ‘cladding’ layers and placed into a binding graphite matrix to form fuel compacts [23]. The inner kernel consisted of either an actinide oxide or carbide, typically uranium (with thorium used for breeding kernels). The earliest microparticle fuel, termed laminar, had only a single pyrolytic carbon coating layer for each particle (see Figure 9). However, this design suffered from issues of poor fission product retention, as the single protective layer was damaged by fission recoils and gas production in the kernel during fuel burnup.

To improve stability during irradiation, a ‘buffer’ layer of low-density carbon was added to surround the fuel kernel and separate the protective PyC layer from the fuel while providing volume for fission gas retention. This particle type was termed BISO, for bi-structural isotropic-type fuel. However, the BISO coated-particle fuel still had difficulty with retaining metallic fission products, so a further design evolution was pursued that consisted of adding a layer of SiC in between two layers of PyC outside the buffer layer. Termed TRISO (for tri-structural isotropic), this final major iteration in coated particle fuel design would serve as the reference fuel configuration for all later coated-particle fueled reactor designs, due to its superior fission product retention performance and irradiation stability. See Figure 9 for a visual representation of the major phases of coated-particle fuel development.

Since the microparticle fuel was self-encapsulated, the metallic fuel cladding present in previous gas-cooled reactor designs could be eliminated. This enabled the use of entirely graphite-based materials in the active core region, which was desirable for high-temperature operation (graphite being capable of operating at bulk temperatures of 1000 C or greater without difficulty) and improved neutron economy (as graphite parasitically absorbs fewer neutrons than most metallic materials). Since graphite can be manufactured in a variety of shapes, this provides considerable freedom in the design of the fuel assemblies for reactors incorporating coated-particle fuel. The Dragon reactor experiment’s assembly geometry consisted of a cluster of seven graphite fuel tubes filled with fuel compacts sandwiched between integral upper and lower axial reflector regions (See Figure 10 and Figure 11), with a total assembly height of 318 cm [22] [45]. The six outer graphite/fuel tubes were easily separable from the innermost, seventh fuel/graphite tube. Thus, this seventh fuel tube could be replaced by an alternative configuration to provide a flexible small-scale fuel and materials testing capability available inside each assembly loaded into the core.

Figure 11. Isometric-view photograph of a Dragon reactor experiment fuel assembly [22]

37 of these assemblies were placed into a triangular core lattice and surrounded by two rings of graphite reflector blocks, the innermost of which is easily removable (See Figure 12). Note that the fuel assemblies tended to have very little pure graphite volume in the actively fueled axial slice, leading to a relatively homogeneous distribution of fuel and moderator in this axial region. However, the fuel assemblies did have large fuel-free upper and lower graphite reflectors, which helped to reduce the neutron leakage and also provided an efficient moderation volume. To further overcome the poor neutron economy of a small, high-leakage test reactor, Dragon used highly-enriched fuel particles, up to 90% U-235 [46].

Figure 12. Top-down view of the Dragon reactor core, showing the 37 fuel assembly positions [22]

2.2.2 Peach Bottom Unit 1

Peach Bottom Unit 1 was the first HTGR constructed and operated in the United States [47]. Intended as a small prototypical HTGR by the plant vendor, General Atomics, Peach Bottom had a thermal power of 115 MWth and an electrical power of 40 MWe. Unlike the Dragon reactor experiment, Peach Bottom was connected to the grid and produced usable electrical power during its operating lifetime, from July 1967 to October 1974. While Dragon pursued a strategy of maximizing the testing capabilities of the reactor experiment such that the design focused on enabling flexibility and incorporated multiple fuel and materials irradiation experiment positions, Peach Bottom was designed to operate much like any other power reactor, albeit at a smaller scale and with the understanding that operational tests would be performed and multiple types of fuel would be incorporated throughout its lifetime. Additional instrumentation was installed to monitor the performance and reliability of the fuel and system components during operation, and a limited capability for test assemblies in a few lattice positions was utilized.

Peach Bottom was designed by General Atomics (GA), and would be the first GA HTGR constructed and operated [47]. The US AEC entered into an agreement with GA in 1958 to share research and development costs for the design of a US HTGR. At the same time, the AEC issued an invitation to US power utilities to build and operate a prototypical HTGR. The Philadelphia Electric Company and several other utilities throughout the US formed a consortium named High

Temperature Reactor Development Associates, Inc. to collaborate with the AEC in the deployment of the Peach Bottom power plant.

Peach Bottom’s first core consisted of 804 fuel elements surrounded by free volume for coolant flow [48]. The second core’s layout and assembly design was essentially the same. The fuel elements were cylindrical graphite cylinders consisting of an outer graphite sleeve protecting annular fuel compact rings stacked on a central graphite spine. The outer diameter of the fuel elements was 8.89 cm (2.5 in) with a total height of 3.658 m (12 ft). The annular ring fuel compacts were 7.62 cm (3 in) tall, with 30 compacts in an assembly, giving an active core height of 2.286 m (7.5 ft). The graphite upper reflector and top spacer ring, located above a top cap that secured the fuel compacts, were 59.44 cm (23.4 in) tall, and the lower reflector (33.02 cm, 13 in), internal integrated fission product trap assembly (32 cm, 12.6 in), and bottom connector (26.42 cm, 10.4 in) were a combined 66.04 cm (26 in) tall. See Figure 13 for a diagram of the nominal Peach Bottom Unit 1 first-core fuel assembly.

Figure 13. Peach Bottom HTGR nominal fuel assembly [48]

The total core height was 3.658 m (same as the assembly height, since the upper and lower reflectors were incorporated into the fuel assemblies), and the total core diameter was 3.962 m (13 ft), including the fixed radial graphite reflector with thickness of 0.610 m (2 ft), giving an active core with radius of 2.74 m (9 ft) and height, as mentioned previously, of 2.29 m (7.5 ft). Peach Bottom also incorporated various burnable absorbers and enrichment zoning to flatten its radial and axial power profiles, which was important because it was intended to operate in a single-batch once-through mode. See Figure 14 for an overview of the Peach Bottom initial core configuration.

Figure 14. Top-down view of Peach Bottom Unit 1 core [48]

While the nominal assembly design did not evolve significantly in the transition from the first core to the second, the microparticle design was improved significantly. Core 1 utilized UC-ThC fuel kernels coated with a single pyrolytic carbon layer, in the ‘laminar’ configuration as incorporated in early cores of the Dragon reactor experiment. As with the Dragon fuel, these

While not designed for extensive reconfiguration and testing like the Dragon reactor experiment, Peach Bottom Unit 1 did test three types of alternate fuel assembly designs during commercial power operation. Two assemblies of the first type tested, named Proof Test Elements (PTE), were irradiated in Core 1 as an experimental proof of concept and test bed for the fuel design proposed for GA’s next HTGR design, the Fort St. Vrain reactor. PTE-1 and PTE-2 had active fuel regions of the same height as the regular fuel assemblies (2.286 m, or 7.5 ft), consisting of four stacked hexagonal graphite blocks with cylindrical holes drilled through. Into 12 of these holes were slotted cylindrical fuel compacts consisting of a graphite matrix filled with fuel microparticles. Seven other channels facilitated coolant flow for heat removal from the assembly. The structure above and below this hexagonal active fuel region was substantially the same as the standard fuel assemblies, with an integrated upper and lower axial reflector and handling structures. See Figure 15 for a diagram of the PTE-type test fuel assembly.

Figure 15. Peach Bottom PTE-type test fuel assembly [48]

In core 2, one PTE-type assembly continued testing after its irradiation in core 1. In addition, two new types of test fuel assembly were evaluated [48]. The first type, termed simply Fuel Test Elements (FTE), consisted of a sleeved cylindrical active fuel region with the same dimensions (height and diameter) as the standard fuel assemblies. The cylindrical fuel compacts containing fuel microparticles employed by the FTEs were arranged in a circular pattern surrounding an

inner test crucible, all inside of the solid graphite block that comprised the active fuel region’s structure. See Figure 16 for a diagram of the FTE-type test fuel assembly.

Figure 16. Peach Bottom FTE-type test fuel assembly [48]

The final type of test fuel assembly irradiated in Peach Bottom Unit 1 was termed Fuel Pin Test Element, or FPTE. The FPTEs were irradiated on behalf of the UK Atomic Energy Authority, which wanted to evaluate potential commercial HTGR assembly designs. These assemblies were cylindrical, with spacer ribs to provide a flow channel between the outer surface of the assembly and the inner surface of the fuel channel in the graphite core structure. Inside the active region of the assembly cylinder, an annular cylindrical fuel compact was sandwiched between outer and inner graphite block structures, with the innermost volume (inside the inner graphite block structure) forming another channel for coolant flow. See Figure 17 for a diagram of the

FPTE-Figure 17. Diagram of the FPTE-type test fuel assembly irradiated in Peach Bottom core 2 [48]

2.2.3 Fort St. Vrain

After designing Peach Bottom Unit 1, GA moved on to the design of a larger commercial power reactor for the Public Service Company of Colorado named Fort St. Vrain [26]. Fort St. Vrain was developed in part with the financial support of the AEC through its Power Reactor Demonstration Program. The 842 MWth/330 MWe helium cooled, graphite-moderated, TRISO-fuel plant was the first large-scale prismatic-block commercial HTGR operated anywhere in the world.

Fort St. Vrain used graphite block fuel assemblies stacked on top of each other to form its active core region. These assemblies were a scaled-up version of the hexagonal PTE-type test assemblies irradiated in Peach Bottom Unit 1. Each assembly measured .79 m (31 in) tall by .36 m (14 in) flat-to-flat [49]. 100 holes drilled through each block served as coolant channels for heat removal, dispersed in a triangular lattice arrangement with 210 channels filled with cylindrical microparticle-fuel-containing compacts. Fort St. Vrain utilized TRISO microparticle fuel, with both fissile (low-enriched uranium carbide) particles and fertile (thorium carbide) particles dispersed in the matrix. These fuel blocks were more robust and easier to refuel than the multi-component assemblies incorporated in the Peach Bottom Unit 1 design. See Figure 18 for a photograph of the FSV fuel block. This fuel block design would subsequently become the baseline design for most future prismatic-block type HTGR designs.