NUC Workshop

Raleigh, North Carolina, USA
September 17-18, 2019
James B. Hunt Jr. Library (1070 Partners Way, Raleigh, NC)
Institute of Emerging Issues Conference Room (4th floor; use elevator)
Centennial Campus

Hosted by North Carolina State University (NCSU)

The advanced reactor system is expected to be a game-changer for nuclear power and energy markets, and helps the transition to a carbon-free future. However, the uncertainties associated with the design, analysis, test, licensing, and construction present major challenges to its deployment. This two-day workshop intends to bring together researchers and stakeholders from nuclear industry, universities, national laboratories, and government to discuss challenges, needs, and emerging trends in innovations that enable acceleration of development, licensing, and commercialization of advanced reactor technologies. The presentations, panel discussions, and breakout will be centered around the integration of the following topics of interest:

Reactor and system designs
Modeling and simulation
Data analytics
Fuel design and manufacturing
Safety and licensing
Economics
Advanced fuel cycles
Operation and control
Education and training

To register, please follow this link: https://go.ncsu.edu/nuc-workshop

Logistics:

Transportation:

Driving direction (to NCSU parking structures): Partners Way, Raleigh NC 27606
Parking permit: Visitors should use paid parking lots (‘paylots”) which have machines which accept credit cards ($10) http://www2.acs.ncsu.edu/trans/parking/index.html#payspace
To avoid parking hurdles, recommend using carpool, taxi or Uber service

Accommodation:

list of hotels can be found at https://admissions.ncsu.edu/visit/accomodations/

For administrative matters: Hermine Kabbendjian (hkabbend@ncsu.edu)

Abdalla Abou Jaoude

Distinguished Postdoctoral Research Associate

Idaho National Laboratory

Title: In-Pile Irradiation Testing for Molten Salt Reactors

Abstract:

Molten Salt Reactors (MSR) have garnered significant interest from the private sector, with over eight companies in North America developing designs. While some concepts rely on previously tested salts (e.g. during the Molten Salt Reactor Experiment), others are proposing new and untested salts. These new mixtures are likely to require irradiation testing prior to licensing. What would such testing look like? What should be prioritized? How would the experiment capsules be designed? The presentation will explore the history of salt irradiation testing, discuss current needs, and start a conversation on how future experiments would be designed.

Bio:

Dr. Abdalla Abou Jaoude is an Advanced Reactor Core Analyst at Idaho National Laboratory (INL). His research focus includes core design and modeling (notably for the Versatile Test Reactor), nuclear economics, and irradiation testing. He is currently leading a feasibility study on the deployment of a molten salt experiment loop within the INL Advanced Test Reactor. Abdalla began his assignment at INL as the deBoisblanc Distinguished Postdoctoral Research Associate. Prior to that, he obtained a Ph.D. in nuclear engineering from the Georgia Institute of Technology. His dissertation was on the evaluation of a mixed-spectrum core configuration that improves the proliferation resistance of long-lived reactors. During his PhD he was awarded the Sam Nunn Security Fellowship, and completed the TI:GER entrepreneurship program. Abdalla also obtained his Masters in Engineering from Imperial College, London.

Vivek Agarwal

Research Scientist

Idaho National Laboratory

Title: Built-in Versus Bolt-on Approach to Support Automation, Operation, and Maintenance Strategy

Maria Avramova

Associate Professor

North Carolina State University

Title: Development and Qualification of State-of-the-Art Subchannel Methods for Advanced Reactors

Abstract:

The current US Department of Energy programs on the development of high-fidelity multi-physics analysis tools, such as NEAMS (Nuclear Energy Advanced Modeling and Simulation) and CASL (Consortium for the Advanced Simulation of Light Water Reactors), have indicated that the subchannel methods are being viewed as efficient and capable of accurately predicting core-wide thermal-hydraulics conditions in acceptable clock time on computer platforms available to the nuclear industry. This talk will discuss the gaps of the “legacy” subchannel methods for advanced reactor modeling and simulations. CTF, the North Carolina State University (NC State) and Oak Ridge National Laboratory (ORNL) jointly developed subchannel code, will be used as an example. CTF is state-of-the-art thermal-hydraulic code, which provides the best available subchannel methods to the scale of an entire core. CTF has been officially part of the DOE CASL project and now is becoming part of the new starting DOE program ModSimX. Although there have been numerous developments (primarily at ORNL) for enabling CTF for solid fuel molten salt reactor modeling, these will not be addressed in this talk. Instead, the focus will be on the ongoing work at NC State to extend the CTF modeling capabilities to Sodium Fast Reactors. This includes incorporating the sodium coolant properties to the code as well as adding correlations for the friction factor, flow mixing coefficient and conduction heat transfer. CTF validation with the Experimental Breeder Reactor-II (EBR-II) Shutdown Heat Removal Tests (SHRT-17) an (SHRT-45R) will be presented.

Bio:

Dr. Maria Avramova is an Associate Professor in Nuclear Engineering at North Carolina State University (NC State). She serves as a Director of the Consortium for Nuclear Power at NC State and Coordinator of the CTF Users’ Group. Dr. Avramova earned her Ph.D. degree in nuclear engineering from the Pennsylvania State University (Penn State) in 2007. Prior to joining Penn State, she held a research scientist position at the Institute of Nuclear Research and Nuclear Energy, Bulgarian Academy of Science, Sofia, Bulgaria. The research interests of Dr. Avramova include nuclear reactor core thermal-hydraulics and design, transient and safety analysis, multi-physics and multi-scale simulations, verification, validation, uncertainty and sensitivity analysis. Over the years, Dr. Avramova has led high visibility international projects such as the Organization for Economic Co-operation and Development, Nuclear Energy Agency (OECD NEA) / United States Nuclear Regulatory Commission (U.S. NRC) BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark and the OECD-NEA/U.S.NRC PWR Subchannel and Bundle Tests (PSBT) benchmark. She is a member of OECD NEA Expert Group on Core Thermal-Hydraulics (EGCTH), Expert Group on Uncertainty Analysis in Modeling (EGUAM), and Expert Group on Multi-Physics Experimental data, Benchmarking and Validation (EGMPEBV).

Paolo Balestra

Nuclear Engineering Research Scientist

Idaho National Laboratory

Title: Pronghorn: A Coarse Mesh Thermal – Hydraulics Application Based on MOOSE for Advanced Reactor Concepts

Abstract:

High temperature gas cooled reactors (HTGRs) are candidates for timely Gen-IV reactor technology deployment because of high technology readiness and walk-away safety. Among HTGRs, pebble bed reactors (PBRs) have attractive features such as low excess reactivity and higher efficiency; on the other hand, they pose unique modeling and simulation challenges to analysts and reactor designers such as long Design Basis Accident (DBA) scenarios, burnup and power distribution depending on pebble recirculation, and radiative heat transfer in the gas-filled gaps. Full-plant reactor safety calculations are needed for proving the safety of the reactor design for licensing purposes; these calculations must be performed by low to intermediate fidelity codes such as system and porous-media based codes. In order to face these challenges and take advantage of the multiphysics object-oriented simulation environment (MOOSE), Idaho National Laboratory (INL) is currently developing a porous-media multi-dimensional, coarse mesh thermal hydraulic application named Pronghorn. Pronghorn solves the compressible Euler equations along with closure equations to simulate the interaction between the solid phase and the fluid phase on mass, momentum, and energy conservation. As a MOOSE based application, Pronghorn is easily maintainable and benefits from the plethora of existing multiphysics models and algorithms in the MOOSE ecosystem. The code is currently undergoing an intensive and systematic verification and validation program to test the code in the widest possible range of relevant scenarios and new advanced features such as tracking of the pebbles burnup and gas mixture treatments will be available soon. The current code performance and results demonstrate the suitability of Pronghorn as an efficient and flexible PBRs modeling tool for designing and testing purposes.

Bio:

Dr. Paolo Balestra obtained his Master’s Degree in energy engineering in 2012 and his Ph.D. in energy and environment with focus on nuclear engineering in the 2017 at the “La Sapienza ” University of Rome. Currently he is working as a nuclear engineering research scientist at the advanced reactors technology department of the Idaho National Laboratory. His research efforts are currently focused on two areas: the analysis and optimization of the new generation of advanced reactor design and the development and validation of dedicated advanced reactor designs simulation tools. During his Ph.D. he accumulated experience as reactor analyst participating to different international projects devoted to the validation of three-dimensional Neutron Kinetic and thermal hydraulic system codes for light water reactors and sodium cooled fast reactors such as and not limited to, the EU FP7 project “JASMIN” (2013-2015), the IAEA Benchmark analysis for the EBR-II Shutdown Heat Removal Tests (2014-2016) or the OECD/NEA Oskarshamn-2 instability benchmark (2013-2016). He spent the last year of his Ph.D. working as multiphysics tools developer at INL within the High-Temperature Gas-Cooled Reactor Methods and Simulation group (2016-2017). During his postdoc at NCSU he was in charge of the maintenance and the release of the thermal hydraulic analysis tool “CTF” and was working on the training database generation for the ARPA-E founded project Nearly Autonomous Management And Control System for advanced reactors (2017-2018). Author of more than 15 papers in peer-reviewed International conferences and 5 publications in international Journals.

Youssef Ballout

Director, Reactor Systems Design & Analysis Division

Idaho National Laboratory

Title: Micro-reactor Program and Current Status

Ronald L. Boring

Distinguished Human Factor Scientist

Idaho National Laboratory

Title 1 : Use of Simulators for Design of Advanced Reactor Control Rooms

Abstract:

Staffing and control are crucial considerations in the development and licensing of advanced reactors. It is incorrect to assume that the control system of a new reactor is a tail-end design activity. The very viability of advanced reactor designs hinges on their ability to be economically competitive with fossil plants for producing electricity, and staffing is key to managing operational costs of future plants. The control system is one of the key factors in staffing levels; as such, it must be an integral part of conceptual reactor design activities. While advanced reactor concepts may prove the harbingers of greater automation in nuclear power, even an automated system must periodically be monitored and calibrated by human operators. Advanced reactors do not solve the challenges of human-in-the-loop control through full automation. Instead, it is necessary to create an automation control environment that enhances human-system interactions. Idaho National Laboratory has developed control room prototyping tools and capabilities to jumpstart control system development for advanced designs. Using simplified plant simulation models and visualization tools, it is possible to prototype control room designs and validate them using operator-in-the-loop evaluations. In this manner, the concept of operation may be developed in parallel with other aspects of reactor design. Along the way, key decisions on the staffing requirements and level of automation may be determined and demonstrated, allowing a clear path to licensing of advanced reactors.

Title 2: Modeling Human Reliability for Advanced Digital and Autonomous Control Rooms

Abstract:

The challenge of new technology is that in many cases it is newer than the tools used to evaluate it. Such is clearly the case with human reliability analysis (HRA). HRA is a framework to identify and quantify the human component of system risk. Originally designed to model human operators in analog control rooms, HRA has not fully kept pace with advances in digital human-system interfaces (HSIs). Digital technologies are being incorporated into control rooms in the form of control room modernization and new builds. These digital HSIs potentially change the types of tasks operators perform (e.g., more monitoring due to increased automation). As a result, human error types and probabilities may be different than for analog control rooms. Still, HRAs for new reactors are being completed with methods that predate digital HSIs. INL developed Rancor, a microworld simulator, as a simplified nuclear power plant control interface, affording collection of human performance data on advanced reactor concepts that have not yet been employed in a full-scope simulator. Collecting such data is the prerequisite for developing and refining HRA for digital HSIs to support the licensing process for new reactors.

Bio:

Ronald Laurids Boring, Ph.D., is a Distinguished Human Factors Scientist at Idaho National Laboratory, where he has led research projects for the U.S. Nuclear Regulatory Commission, NASA, the U.S. Department of Energy, the Canadian Nuclear Safety Commission, the Department of Defense, and the Norwegian Research Council. He previously worked as a human reliability researcher at Sandia National Laboratories, a usability engineer for Microsoft Corporation and Expedia Corporation, a guest researcher in human-computer interaction at the National Research Council of Canada, and a visiting human factors scientist at OECD Halden Reactor Project. He and his research team developed the Guideline for Operator Nuclear Usability and Knowledge Elicitation (GONUKE) for conducting human factors in support of nuclear technologies, the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER) dynamic human reliability framework, and the Advanced Nuclear Interface Modeling Environment (ANIME) for prototyping digital interfaces in nuclear power environments. Dr. Boring is the founder of the Human Systems Simulation Laboratory.

Dr. Boring has a Ph.D. in Cognitive Science from Carleton University. He was a Fulbright Academic Scholar to the University of Heidelberg, Germany. He has published over 250 research articles in a wide variety of human reliability, human factors, and human-system interaction forums. He is the founder and chair of the Human Error, Reliability, Resilience, and Performance conference, he is co-chair for the 2019 American Nuclear Society Nuclear Power Instrumentation, Controls and Human-Machine Interface Technology (ANS NPIC&HMIT) conference, and he is Chair for the 2019 Annual Meeting of the Human Factors and Ergonomics Society.

Jacopo Buongiorno

TEPCO Professor

Massachusetts Institute of Technology

Title: New Nuclear Needs a Paradigm Inversion

Abstract:

The nuclear industry, as presently configured, is inherently dis-advantaged with respect to most other industrial sectors: it relies on large and costly machines, delivered by an inefficient construction sector, requiring a lengthy safety-driven licensing process, and producing grid-connect electrons (a near-zero-margin commodity). If advanced nuclear technology is to thrive in the 21st century, its development, demonstration and deployment (DD&D) paradigm must be completely reversed. In this talk we will analyze what that entails and the massive opportunities it may create.

Bio:

Jacopo Buongiorno is the TEPCO Professor of Nuclear Science and Engineering at the Massachusetts Institute of Technology (MIT), and the Director of Science and Technology of the MIT Nuclear Reactor Laboratory. He teaches a variety of undergraduate and graduate courses in thermo-fluids engineering and nuclear reactor engineering. Jacopo has published 90 journal articles in the areas of reactor safety and design, two-phase flow and heat transfer, and nanofluid technology. For his research work and his teaching at MIT he won several awards, among which the ANS Outstanding Teacher Award (2019), the MIT MacVicar Faculty Fellowship (2014), the ANS Landis Young Member Engineering Achievement Award (2011), the ASME Heat Transfer Best Paper Award (2008), and the ANS Mark Mills Award (2001). Jacopo is the Director of the Center for Advanced Nuclear Energy Systems (CANES). In 2016-2018 he led the MIT study on the Future of Nuclear Energy in a Carbon-Constrained World. Jacopo is a consultant for the nuclear industry in the area of reactor thermal-hydraulics, and a member of the Accrediting Board of the National Academy of Nuclear Training. He is also a member of the Naval Studies Board, a Fellow of the American Nuclear Society (including service on its Special Committee on Fukushima in 2011-2012), a member of the American Society of Mechanical Engineers, and a participant in the Defense Science Study Group (2014-2015).

Sunil Chirayath

Associate Professor

Texas A&M University

Title: Progress in the Safeguards Approaches for the Molten Salt Reactors

Abstract:

Several advanced nuclear reactor designs and corresponding R&D are progressing, which are Generation IV-type reactors. The four attributes of a Generation IV reactor are: highly economical, enhanced safety, proliferation resistance, and reduced amount of nuclear waste compared to the Generation II, III, III+ reactors. Among them, prominent reactor designs are the molten salt reactors (MSRs), specifically (a) the fluoride-salt-cooled high-temperature reactor (FHR) that uses pebble-type solid fuel and cooled by molten salt, and (b) salt-fueled reactors that used dissolved fuel in the molten salt. There is a need for imminent R&D effort on the nuclear safeguards approaches development for these reactor types. A project is in progress at Texas A&M University in collaboration with the Tokyo Institute of Technology to develop nuclear material balance areas (MBAs), nuclear material key measurement points (KMPs) and nuclear material balance periods (MBPs) for Advanced Reactor Fuel Cycles for supporting special nuclear material accounting and monitoring to guard against nuclear proliferation and malicious acts, which also considers MSRs. Progress made on this project will be discussed during the deliberations of the workshop.

Bio:

Dr. Sunil Chirayath is an Associate Professor of Nuclear Engineering at Texas A&M University (TAMU). He is also the Director of the Center for Nuclear Security Science and Policy Initiatives (NSSPI) at TAMU since 2015. Dr. Chirayath’s expertise is in nuclear security and safeguards applied to nuclear fuel cycle facilities and in Monte Carlo Radiation Transport. Prior to his start at Texas A&M in 2007, he served the Indian Atomic Energy Regulatory Board (AERB) in various capacities for 19 years in safety review of nuclear power plants. He has over 170 technical publications in refereed journals (46) and national/international conference proceedings (100+).

Mark DeHart

Distinguished Staff Engineer

Idaho National Laboratory

Title: Multi-Physics Simulation of Advanced Reactor Designs using MOOSE-Based Reactor Physics Methods

Abstract:

This presentation will give an overview of the MAMMOTH Reactor Physics Tool developed at Idaho National Laboratory, and application to transient simulation for TREAT. New members of the MOOSE ecosystem, including Sockeye, DireWolf and Velociraptor, and their application toward micro reactor systems will be discussed.

Bio:

Mark DeHart is a Directorate Fellow at Idaho National Laboratory. Dr. DeHart leads a team of reactor physicists and computational methods staff performing applied multi-physics methods for numerous reactor types, initially supporting the Advanced Test Reactor (ATR), with a focus since 2015 on supporting advanced methods for Transient Test Reactor (TREAT) transient simulations for both core and experiment. He currently leads an LDRD that seeks to couple the Naval Nuclear Laboratory (NNL) Monte Carlo code MC21 to the MOOSE-based BISON fuel performance code. His team’s current work focuses on advanced reactor designs, supporting both the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program and the US NRC. He is also engaged with micro-reactor methods support with Westinghouse as part of the DOE MEITNER Resource Team. Dr. DeHart has extensive experience in reactor physics, criticality safety, depletion and spent fuel characterization, cross-section processing, and computer code verification and validation. He holds BS, MS, and Ph.D. degrees in nuclear engineering from Texas A&M University. He is a Fellow of the American Nuclear Society (ANS) and a member and past Chair of the local section of ANS. He has more than 150 publications in journals, conference proceedings, and national laboratory reports related to computational methods and applications in reactor physics, radiation transport, criticality safety, validation and depletion methods for spent nuclear fuel.

Charles Forsberg

Research Scientist

Massachusetts Institute of Technology

Title: Nuclear Reactors with Heat Storage to Boost Revenue and Replace Fossil Fuels

Abstract:

The electricity grid is changing because of (1) the addition of wind and solar that creates volatile electricity prices including times of zero-priced electricity and (2) the goal of a low-carbon world that requires replacing fossil fuels that provide (a) energy, (b) stored energy and (c) dispatchable energy. Wind and solar provide energy but not the other two other required energy functions that are provided fossil fuels. Nuclear energy with heat storage can fully replace fossil fuels including providing dispatchable electricity. Heat storage enables reactors to maximize revenue by selling electricity at times of high prices and buying electricity that is converted to stored heat at times of low or negative electricity prices. These systems enable a base-load nuclear plant to provide peak electricity production that is significantly above the base-load electricity production capacity. These options were examined at the July 23-24, 2019 MIT/INL/Exelon workshop: Heat Storage for Gen IV Reactors for Variable Electricity from Base-load Reactors: Changing Markets, Technology, Nuclear-Renewables Integration and Synergisms with Solar Thermal Power Systems. The results from this workshop are described that include (1) the institutional requirements, (2) examination of heat storage systems capable of gigawatt-hour heat storage and (3) the path forward. The changes in the market may make traditional base-load nuclear reactors uneconomic. The most economic (profitable) reactors may be those that best integrate the reactor with heat storage and the power cycle because of the potential to double total plant revenue.

Bio:

Dr. Charles Forsberg research areas include Fluoride-salt-cooled High-Temperature Reactors (FHRs) and utility-scale heat storage including Firebrick Resistance-Heated Energy Storage (FIRES). He teaches at MIT the fuel cycle and nuclear chemical engineering classes. Before joining MIT, he was a Corporate Fellow at Oak Ridge National Laboratory. He is a Fellow of the American Nuclear Society, a Fellow of the American Association for the Advancement of Science, and recipient of the 2005 Robert E. Wilson Award from the American Institute of Chemical Engineers for outstanding chemical engineering contributions to nuclear energy, including his work in waste management, hydrogen production and nuclear-renewable energy futures. He received the American Nuclear Society special award for innovative nuclear reactor design and is a Director of the ANS. Dr. Forsberg earned his bachelor’s degree in chemical engineering from the University of Minnesota and his doctorate in Nuclear Engineering from MIT. He has been awarded 12 patents and published over 300 papers.

Kevin Han

Assistant Professor

North Carolina State University

Title: Integrated Approach to Design and Modular Construction through Advances in Visual Data Analytics, BIM, and Robotics

Abstract:

The nuclear construction industry has experienced huge escalations in overnight construction and schedule delays due to (i) extremely stringent nuclear safety and quality assurance (QA) standards, (ii) inexperience in managing and staffing these standards, (iii) excessive paperwork, and (iv) supply chain delays due to rework. This presentation will show our integrated approach to design and construction to lower fabrication and overnight construction cost and time. Our work includes the established pipeline that integrates various advanced modeling and simulation (M&S) tools with Building Information Modeling (BIM) to deal with design changes during construction. The presentation will also include the development of Construction Performance Modeling and Simulation (CPMS) to digitally manage QA/Quality Control (QC) inspections and construction progress through the use of reality capture technology and BIM. The presented solution will be embedded into the supply chain loop to ensure ongoing quality control, simulation of weekly progress and work schedules, and timely decision support throughout construction.

Bio:

Kevin Han is an Assistant Professor in the Department of Civil, Construction, and Environmental Engineering and the Center for Nuclear Energy Facilities and Structures at NC State University. He received a PhD in Civil Engineering and Master of Computer Science at the University of Illinois at Urbana-Champaign, an MS in Civil Engineering, a BA in Architecture, and Minor in Structural Engineering at the UC Berkeley. His research focuses on automating construction engineering and management through advances in robotics, visual data analytics, and building information modeling. His work has been recognized through best paper and poster awards at ASCE academic conferences.

Peter Hastings

Vice President – Regulatory Affairs and Quality

Kairos Power

Title: Kairos Power Perspectives on Advanced Reactor Innovations

Florent Heidet

Manager – Advanced Nuclear Energy Systems

Argonne National Laboratory

Title: The Versatile Test Reactor, irradiation testing capabilities for advanced reactors

Abstract:

The Versatile Test Reactor (VTR) is being designed to allow accelerated irradiation testing to support the development of most advanced reactor technologies. It will enable a variety of coolants to be used in the testing channels in order to provide close to prototypic conditions for fuel and material testing. This will be achieved through the use of enclosed capsules as well as through the use of cartridge loops. With a peak flux in excess of 6×1015 n/cm2-s, the VTR will offer damage rates as high as 65 dpa/year, as well as over 10 concurrent testing locations with large usable volumes. We will present the current state of the project, an overview of the core design, and some of the on-going activities.

Bio:

Florent Heidet is the manager of the Advanced Nuclear Energy Systems group at Argonne National Laboratory. Florent is an expert on advanced reactors, with over 10 years of experience designing and analyzing various Gen-IV reactors. He is responsible for the Versatile Test Reactor core design activities, leading the multi-disciplinary team responsible for delivering the VTR core. Florent has been the lead core designer for the Prototype Gen-IV SFR project between KAERI and Argonne, has initiated and led the Molten Salt Reactor modeling & simulation and analysis activities at Argonne, and is currently the Argonne technical lead for the Transformational Challenge Reactor program. His interests are wide and include everything that allows enabling construction of advanced reactors. He graduated with a Ph.D. from UC Berkeley in nuclear engineering and a M.Sc. from the ENSAM ParisTech in Mechanical Engineering.

Jason Hou

Assistant Professor

North Carolina State University

Title: Multi-Objective Core Optimization Framework for Advanced Reactors

Abstract:

The optimization of a Sodium-cooled Fast Reactor (SFR) core is a challenging process, due to the large number of design parameters, the nonlinearities among inputs and outputs, and the complicated correlation among output parameters. This presentation shows a generalized framework for the SFR core optimization by coupling the sensitivity analysis, advanced optimization algorithm, and optionally the surrogate modeling. The genetic algorithm was selected as the optimization method for its robustness, while the option of surrogate modeling was also explored to alleviate the computational burden caused by employing the direct core physics simulation and thus enhance the efficiency of the optimization. The normalized deviations of performance parameters of the near optimal solution from those of the ideal core were calculated and used as criteria to down select the final core design. The developed framework was applied to the Advanced Burner Test Reactor (ABTR) core, and optimal solutions were determined by balancing various objectives simultaneously, including peak fast flux, core volume, power, reactivity swing, plutonium mass feed, while at the same time satisfying the predefined constraints due to safety or economics considerations. The optimal ABTR core design obtained using the direct physical simulation and surrogate model were compared and discussed. It was found that using the accurately constructed surrogate models can significantly reduce the required computational time while maintaining satisfactory accuracy.

Bio:

Dr. Hou is an Assistant Professor of Nuclear Engineering at North Carolina State University. His expertise of research includes the multi-physics reactor simulation, advanced reactor and system design, core loading and fuel cycle optimization, sensitivity and uncertainty (S/U) analysis in modeling of various reactor systems, high-fidelity reactor core simulator, and hybrid Monte Carlo (MC) and deterministic method. He serves as the co-coordinator of two NEA/OECD benchmarks: uncertainty analysis in the modeling of LWR benchmark and the homogenization-free time-dependent neutron transport benchmark.

Hou received his B.S. degree in Engineering Physics from Tsinghua University, China. He holds M.S. and Ph.D. degrees in Nuclear Engineering from University of Michigan and Pennsylvania State University, respectively. Prior to joining the NC State faculty, he was a postdoctoral scholar in the Department of Nuclear Engineering at the University of California, Berkeley. He held a position of Research Assistant Professor in the Department of Nuclear Engineering at NC State University.

John Link

Zachry Nuclear Engineering

Principal Consultant

Title: Applications of GOTHIC for Simulation of Advanced Reactors

Abstract:

GOTHIC™ is a versatile, general purpose thermal-hydraulics tool that includes both system level and CFD attributes. The software provides an integrated analysis environment with 1) a graphical user interface (GUI) that allows for fast, flexible creation or modification of models, 2) a numerical solver that includes parallel processing capabilities, and 3) a post-processor for quickly evaluating simulation results. It solves the conservation equations for mass, momentum, and energy for multicomponent, multi-phase flow in lumped parameter and multi-dimensional geometries (1, 2, or full 3D), including the effects of turbulence, diffusion and buoyancy. The flexible nodalization options using a domain decomposition approach allows GOTHIC to provide computationally efficient solutions for design, optimization and safety analysis.

While GOTHIC has traditionally been considered as a containment analysis tool, the capabilities have continuously evolved over 30+ years of active development under a nuclear quality assurance (NQA) program that complies with 10 CFR 50 Appendix B, Part 21 and the applicable portions of ASME NQA-1. The fundamental physical models, generalized fluid property framework and other capabilities now available in GOTHIC make the software applicable for simulating many of the advanced, non-LWR concepts under development. Fluid property tables can be generated using data from NIST RefProp or using a stand-alone program to evaluate the Equation of State (EOS). GOTHIC 8.3(QA), which represents the latest release of the software, includes properties for light water, heavy water, sodium (Na), sodium-potassium (NaK), lead (Pb), lead-bismuth eutectic (LBE) and six molten salts (NaCl-MgCl2, LiF-BeF2, LiF-NaF-KF, NaF-ZrF4, KF-ZrF4 and NaBF4-NaF).

GOTHIC has been used for design & licensing of existing plants, small modular reactors (SMR) and next generation plant designs. This presentation will summarize relevant attributes, validation benchmarks to experimental data, including EBR-II and MSRE, as well as experience applying GOTHIC for advanced reactors, including TerraPower’s sodium and molten salt reactor concepts.

Bio:

As a Principal Consultant with almost 40 years of nuclear power experience, Mr. Link has a broad range of experience in engineering, training and operations disciplines. His recent work has focused on advanced reactor applications of GOTHIC, including sodium cooled designs such as EBR-II and TerraPower’s TWR as well as molten salt reactors such as the MSRE. He is also the lead GOTHIC trainer, developing and delivering several training classes per year. Prior to joining Zachry, Mr. Link worked as a Shift Technical Advisor and safety analyst at TMI-I, as a containment and safety analyst for AREVA NP and was a U.S. Navy Nuclear Engineering Officer, serving on two nuclear powered cruisers. He also has experience in model development, analyses and review in various thermal hydraulics codes, such as RELAP5, RETRAN, GOTHIC, CONTEMPT, and MAAP.

Patrick McDaniel

Research Professor

University of New Mexico

Title: Why Fast Reactors with Air-Brayton Power Conversion Cycles

Abstract:

Nuclear power in the United States is in a crisis at this time. This talk will discuss several requirements of advanced reactors and power conversion systems that can lead advanced nuclear power systems out of this crisis and become better able to deal with the future demands of the electrical power grid in the United States. The systems to be emphasized will be fast reactors and Air-Brayton power conversion systems. Model results will be presented that demonstrate these solutions can address a future electrical grid with a high penetration of renewable energy systems, reduce the requirements for additional fresh water cooling, and have a minimizing impact on the generation of additional radioactive waste. Safety implications will also be addressed.

Bio:

Patrick McDaniel is currently an adjunct and research professor at the Department of Nuclear Engineering, University of New Mexico. He began his career as a pilot and maintenance officer in the USAF. After leaving the Air Force and obtaining his doctorate at Purdue University, he worked at Sandia National Laboratories in fast reactor safety, integral cross-section measurements, nuclear weapons vulnerability, space nuclear power, and nuclear propulsion. He left Sandia to become the technical leader for Phillips Laboratory’s (became part of Air Force Research Laboratory) Satellite Assessment Center. After 10 years at PL/AFRL, he returned to Sandia to lead and manage DARPA’s Stimulated Isomer Energy Release Project. While at Sandia, he worked on the Yucca Mountain Project and DARPA’s Classified UER-X Program. Having taught at the University of New Mexico in the Graduate Nuclear Engineering Program for 25 years, when he retired from Sandia in early 2009, he joined the faculty at the University of New Mexico full time. He holds a BS degree in Engineering Science from the USAF Academy, an MS in Mechanical Engineering (nuclear option) from CalTech, a PhD in Nuclear Engineering from Purdue University, and an MS in Resource Management from the Industrial College of the Armed Forces.

Eben Mulder

Chief Nuclear Officer

X-Energy

Title: Xe-100 – Aspects of Design Important to its Safety Considerations

Abstract:

This presentation summarizes the design basis of the Xe-100, key fuel performance factors, those factors directly impacting fuel temperature and X-energy’s approach to calculating the source term in both normal and upset conditions. A broad overview will be provided of the way that the in-house source term code, XSTERM interfaces with the design codes.

Bio:

As Senior Vice President and Chief Nuclear Officer at X-energy, Eben works to maintain an architectural framework for the X-energy nuclear program in guiding design and implementation. He is also responsible for the R&D and technology roadmap associated with the development of the program and participates in the business development process. Prior to joining X-energy, Eben served as Chief Scientific Officer for South Africa’s PBMR pebble bed project, and as Corporate Consultant to ESKOM, South Africa’s state utility. He has consulted for several organizations including the U.S. Department of Energy, Vattenfall in Sweden, and the IAEA in Austria. Eben also worked in Germany on the AVR, an experimental pebble bed reactor designed and operated to develop the pebble/spherical fuel elements. Eben holds degrees in Mathematics from the University of Port Elizabeth, a degree in Applied Mathematics from the University of Pretoria and a Doctorate in Nuclear Engineering from the Technical University of Aachen. He is the author of more than 100 works on pebble bed projects and served as Director of the Post Graduate School of Nuclear Sciences and Engineering, North-West University, and also as Extraordinary Professor within the Faculty of Engineering, North- West University.

David Pointer

Group Leader – Advanced Reactor Engineering

Oak Ridge National Laboratory

Title: Building Digital Laboratories for Nuclear Energy Innovation

Abstract:

The rise of readily available high performance computing resources has thrown open the doors to a new era of engineering analysis. In the early 2000s, the birth of contemporary commercial HPC architectures enabled the development of multi-physics frameworks that in turn enabled split-operator integration of multiple physics or engineering codes. Each of these independent solvers may be carefully constructed to address specific characteristics of the equations describing the particular physical phenomena addressed by that code. Significant work in the past decade has focused on refinement of these Picard integration schemes and investigation of alternative Jacobian free methods. Today, petascale computational resources are enabling a new, orthogonal layer of integration in which high-resolution simulations describing a particular physical phenomenon are used to improve lower resolution engineering models through both calibration and direct data methods. These capabilities are necessary to enable the digital optimization of new design concepts, but they are not sufficient alone to serve as digital laboratories for testing new ideas.

True digital twins can serve as digital laboratories for the testing and optimization of new ideas. However, we must build upon the foundation of the multiphysics, multiscale reactor core simulation capabilities developed within several programs supported by the U.S. Department of Energy and others to enable a broader analysis and optimization capability that reaches to the three values that ultimately determine whether a new reactor project is pursued – safety, reliability, and cost.

Bio:

David Pointer (PhD, NE, University of Tennessee, 2001; MS, NE, University of Tennessee, 2000) is the Group Leader for Advanced Reactor Engineering and a Distinguished R&D Staff Member at Oak Ridge National Laboratory (ORNL), where he specializes in computational fluid dynamics (CFD) and heat transfer, experimental fluid dynamics and heat transfer, and nuclear reactor safety. He currently serves as the Deputy Focus Area Lead for Thermal Hydraulics Methods in the Consortium for Advanced Simulation of Light Water Reactors (CASL), which is developing, validating and demonstrating advanced CFD based capabilities for evaluation of thermal hydraulic performance, DNB/CHF limits and crud growth in LWR fuel assemblies. He previously served as the Technical and Program Coordinator for the development of U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Toolkit for advanced reactor and fuel performance and safety analyses.

At ORNL, David contributes to a number of other initiatives including the development of high fidelity experiments to qualify thermal hydraulic and CFD software for the analysis of advanced nuclear power systems; the validation and development of multi-dimensional simulation capabilities for liquid metal, molten salt, high temperature gas and boiling water flows; and the application of CFD methods to a variety of engineering problems in the areas of electric power generation, engine design and vehicle aerodynamics.

Dr. Pointer was the recipient of the 2014 American Nuclear Society Young Member Advancement Award, the 2012 American Nuclear Society Landis Young Member Engineering Achievement Award and the 2007 American Nuclear Society Young Member Excellence Award. He is an elected At-Large Member or the American Nuclear Society Board of Directors, the current Chair of the ANS Thermal Hydraulics Division and a Past-President of the North American Young Generation in Nuclear (NAYGN).

Piyush Sabharwall

Research Scientist

Idaho National Laboratory

Title: Advanced Reactor Thermal Hydraulics Experiments and Modeling

Rachel Slaybaugh

Program Director

ARPA-E

Title: Nuclear and ARPA-E: Activities and Opportunities

Abstract:

ARPA-E has funded a variety of technologies to enable the development and deployment of advanced reactors, and we’re not finished yet. This talk will highlight what we’ve done so far and what we’re thinking about doing next. In particular, we’re interested to discuss how to build collaborations outside of the nuclear industry to take advantage of solutions that have been developed and leverage them.

Bio:

Dr. Rachel Slaybaugh currently serves as a Program Director at the Advanced Research Projects Agency-Energy (ARPA-E). Her focus at ARPA-E includes a wide range of technologies to enable advanced nuclear reactor systems.

Slaybaugh is also an Assistant Professor at the University of California, Berkeley. There, she researches numerical methods for neutral particle transport with an emphasis on supercomputing and advanced architectures. Her research applies to reactor design, shielding, and nuclear security and non-proliferation. Slaybaugh also runs the Nuclear Innovation Bootcamp to train the next generation of people working on nuclear energy. She was appointed to the Nuclear Energy Advisory Committee and she serves as a Senior Fellow at the Breakthrough Institute.

Slaybaugh earned an M.S. and Ph.D. in nuclear engineering and engineering physics from the University of Wisconsin-Madison. She holds a B.S. in nuclear engineering from Pennsylvania State University.

Samantha Smelley

Zachry Nuclear Engineering

Nuclear Engineer

Title: Applications of GOTHIC for Simulation of Advanced Reactors

Abstract:

GOTHIC™ is a versatile, general purpose thermal-hydraulics tool that includes both system level and CFD attributes. The software provides an integrated analysis environment with 1) a graphical user interface (GUI) that allows for fast, flexible creation or modification of models, 2) a numerical solver that includes parallel processing capabilities and 3) a post-processor for quickly evaluating simulation results. It solves the conservation equations for mass, momentum and energy for multicomponent, multi-phase flow in lumped parameter and multi-dimensional geometries (1, 2, or full 3D), including the effects of turbulence, diffusion and buoyancy. The flexible nodalization options using a domain decomposition approach allows GOTHIC to provide computationally efficient solutions for design, optimization and safety analysis.

Bio:

Ms. Smelley is a nuclear engineer with knowledge of advanced reactor design concepts and experience in radiological and dose analyses. During her tenure with Zachry, Ms. Smelley has established expertise in applying GOTHIC for EQ, room heat-up/cool-down, HELB, heat load benchmarking, ELAP/FLEX analyses, and fission product transport. Ms. Smelley is also experienced in performing various onsite/offsite dose consequences analyses, mission dose analyses, shine dose analyses, and shielding determinations for a wide range of clients.

John Henry Sullivan

Consultant, SMR Engineering

Tennessee Valley Authority

Title: TVA Clinch River SMR Project – The PPE Approach to ESPA and Emergency Planning

Abstract:

New reactor siting based on advance technologies under development could benefit from the development of a bounding plant parameter envelope that supports early site permits and early submittal of regulatory exemptions.

Bio:

Forty years of experience in engineering, construction, startup, operation, licensing, and project management in hydro, fossil-fueled, and nuclear power plants. BSNE & BSME NCSU, SRO license, and Professional Engineer.

Currently supporting TVAs Clinch River Nuclear Project, Early Site Permit Application, Design Certification Reviews, and Combined Operating License support including engineering processes, plant parameter envelope, site selection, emergency planning, accident analysis, and SMR design reviews.

Pavel Tsvetkov

Associate Professor

Texas A&M University

Title: Advanced Optimization Methods and Performance Evaluation in Advanced Reactor Design

Abstract:

The emerging novel reactor designs target expanding application areas in a wide variety of operational domains, from electricity to process heat to emergency response to space environments. Furthermore, advanced designs aim at taking full advantage of modern manufacturing technologies going as far as 3D printing of components and entire systems. Ultimately, advanced reactor designs are highly integrated systems. The idea is to develop and deploy a successfully marketable reactor technology allowing to harvest nuclear energy in a system that would have streamlined simplified deployment pathways and operational characteristics reducing the cost of its manufacturing and subsequent lifecycle management from cradle to grave while assuring nearly absolute reliability. The integration and advanced manufacturing, when considered together, suggest new optimization methods yielding desired performance characteristics of advanced reactor systems right out of the box with minimal installation needs. The presentation will focus on the design and optimization methods as well as design space parametric reduction approaches needed to support the integrated robust energy system development and deployment efforts. It will account for methods which are capable to support such system considerations as security and autonomy of their operation. The idea is to develop not only integrated designs but integrated design-development-manufacturing technologies that will yield cost-effective nuclear reactor designs that would be customizable and scalable to the needed applications while being streamlined and simplified to support factory-stage certification and minimized deployment needs. Synergistically, such methods account for not only unit performance but also for manufacturing considerations, fuel cycle and decommissioning. These considerations and approaches yield dramatically reduced uncertainties and risks challenging successful deployment expectations, such as licensing, construction, operational reliability and maintenance. The novel design methods will lead to game-changing innovation towards accelerated commercialization.

Bio:

Dr. Tsvetkov is an Associate Professor in the Department of Nuclear Engineering, Texas A&M University. He has a Ph.D. in Nuclear Engineering from Texas A&M University and M.S. in Nuclear Engineering from Moscow State Engineering Physics Institute (TU), Russia. His research program is focused on novel energy systems meeting global growing needs in sustainable resources. Advanced system design and optimization methods for complex engineered systems enable development of novel sustainable nuclear energy technologies towards “environmentally benign” systems and their deployment for a broad range of applications. Dr. Tsvetkov is a member of ANS, ASME, ASEE, Alpha Nu Sigma and Phi Kappa Phi. At A&M, he serves as the graduate program advisor for Nuclear Engineering as well as Director of the Systems Engineering Program. He authored and co-authored numerous books and book chapters and has over 300 publications in the areas of advanced reactor design methods, analysis and development.