8th International Conference on Simulation Methods in Nuclear Science and Engineering
Abstracts and Biographies
Professor, School of Nuclear Science and Technology, Xi'an Jiaotong University (XJTU) China.
Development of a High Fidelity Numerical Reactor Model and its Uncertainty Quantification
This plenary talk will introduce the current status of a newly developed high-fidelity numerical reactor model by coupling the deterministic neutronics code NECP-X and the subchannel code COBRA-TF. The uncertainty quantification of the model will be also discussed. The NECP-X code was developed by the Nuclear Engineering Computational Physics (NECP) lab at Xi’an Jiaotong University. In this code, a 69-group cross section library and a continuous energy cross section library are included, in which the continuous-energy cross section library is for the resonance self-shielding calculation. A new global-local resonance method based on the pseudo-resonant-nuclide subgroup method is proposed to improve the accuracy of traditional subgroup method. The 2D/1D coupled solver was employed for neutron transport calculation. To overcome the stability issue in the 2D/1D fusion transport method, an improved leakage splitting method was developed and implemented in the code. Furthermore, the pin-resolved-level thermal-hydraulics feedback is given by the sub-channel code COBRA-TF. The VERA progression problems 1 through 7 are tested to verify the coupled model, and the numerical results show that the coupled model agrees very well with the references. Pin-resolved-level N/TH coupling could reduce the maximum fuel temperature by over 10K for the 3D assembly cases compared to the pin-level N/TH coupling calculation. The uncertainty quantification has been performed against the C5G7-TD benchmark. Some preliminary results on the UQ will be presented and discussed.
Prof. Liangzhi Cao received his PhD in 2005 from Xi’an Jiaotong University (XJTU), became a full professor of XJTU in 2014, and gained the Outstanding Youth Fund of Natural Science Foundation of China in 2015. He is the vice Chair of Youth Work of the Chinese Nuclear Society (CNS-YGN), the member of American Nuclear Society (ANS) and a Program Committee member of Reactor Physics Division (RPD). He also serves as the Guest Editor and Editorial Committee Member of Nuclear Engineering and Design (NED), and Editorial Committee Member of Annals of Nuclear Energy (ANE). Prof. Cao is well known for his work in the field of reactor physics, including the resonance self-shielding methods, neutron-transport methods, sensitivity and uncertainty analysis, and code verification and validation, etc. He has authored more than 200 journal papers, co-authored two books in reactor physics analysis methods. He is now leading some key research projects funded by Minister of Science and Technology of China and Natural Science Foundation of China. Prof. Cao is an energetic and creative professor who supervised more than 20 PhD and Master students in reactor physics, most of them are playing key roles in nuclear industry and academy in China.
Professor, Royal Military College (RMC), Canada
Margin Improvement Initiatives: Realistic Approaches
As CANDU reactor cores age, safety margins become particularly tight. Realistic and practical approaches are proposed here to recover lost margins and to support the development of advanced fuel. Research activities at the Royal Military College of Canada, with a focus on the following three topics, will be presented:
1. The use of a small amount of neutron absorbers in fuel bundles to mitigate the fuelling transients and end-flux peaking,
2. The statistical analysis of actual fuel manufacturing data to demonstrate safety margins, and
3. A statistical approach to establish fuel reliability.
Temporary power derating can result from the excess reactivity introduced due to the insertion of fresh fuel bundles into the reactor core. The first project shows that small amounts and appropriate mixtures of neutron absorbers could eliminate xenon-free effect and power ripples, to improve the reactor’s operating margins, with minimal impact on burnup. Details of this project and the use of MCNP, WIMS-AECL and RFSP are presented.
Limit of operating envelope (LOE) approach has little consideration given to the high quality, maturity and precision of current manufacturing processes. Probability distributions were fitted to actual fuel manufacturing datasets provided by a fuel manufacturer. Using the Monte Carlo simulations, the distribution of fuel response was generated from ELESTRES and ELOCA. The second project indicates that the fuel response distributions were far below from the failure limits.
The third project is to develop and conduct a reliability analysis by incorporating manufacturing data, operating histories and models that could be used to determine the probability of fuel failure. The recently developed Fuel and Sheath Modelling Tool (FAST) will be utilized. RFSP will be used to supplement the normal operating data of CANDU cores where data is not readily available. The end-state of the study will yield an approximation of the safety margin that is more accuracy than the LOE method.
The message from this presentation is that the CANDU fuel is indeed flexible, safe and reliable.
Dr. Paul Chan has been working in the CANDU Industry for 30 years. Paul is an expert on nuclear fuel. Numerous industry leadership roles were taken, while Paul was working for AECL (both at the Chalk River and Sheridan Park locations) and Bruce Power.
Paul joined the Royal Military College of Canada, as a Professor and the Manager of the SLOWPOKE-2 Facility, in November 2010. He has trained over 25 graduate students (PhD and Masters) with over 40 journal publications for the last eight years. Paul’s research interests are fuel performance, modeling, safety & licensing and SMR development. Paul currently has nine members in his research group. He collaborates extensively with many organizations and universities such as U of Waterloo, KTH (Sweden), CNL, COG, IAEA and KAERI.
Dr. Chan chaired many international conferences. He is also the Fuel Technologies Division Chair for the Canadian Nuclear Society.
Paul is currently on a sabbatical leave with the IAEA, in Vienna. He is coming back to attend this conference.
Professor and Head of Department of Nuclear Engineering, North Carolina State University (NCSU), United States
Benchmark Approach for Consistent Testing of Multi-Physics Uncertainty Propagation in Reactor Calculations
The objective of the presented benchmark approach is to provide a consistent general framework for multi-scale and multi-physics computation focused on uncertainty propagation and analysis. The full chain from multi-group lattice physics up to full plant transient calculations is considered. A step-by-step approach, fully consistent, is proposed. Compared to previous benchmark approaches the Uncertainty Analysis in Modeling (UAM) approach is the most comprehensive and consistent procedure regarding uncertainty analysis in nuclear engineering. It is organized in three main Phases with several Exercises for each phase, and several test cases for each exercise. The test cases are based on real plant data for the reactor designs of interest and relevant experimental conditions for validation. Special attention will be devoted to interactions between different physics phenomena present in a reactor core (thermal-hydraulics, reactor physics and fuel modelling) in uncertainty quantification and propagation. The corresponding interrelations in uncertainty quantification and propagation as well as their treatment will be shown on the example of some applications. The focus will be on consistency in uncertainty assessment between fine models implemented in fuel performance codes and the rather simplified models implemented in thermal-hydraulics codes, to be used for coupling with neutronics tools. Similarly, the uncertainty quantification on thermal-hydraulic models is established on a relatively small scale, while these results will be used in multi-physics calculations at the core scale, sometimes with different codes. In summary, in addition to multi-physics coupling the uncertainties are propagated through multi-scales of different physics using high-to-low fidelity model information with uncertainties.
Dr. Ivanov received an MEng in nuclear engineering from the Moscow Institute of Power Engineering, Russia. He received his Ph.D. in reactor physics from the Institute of Nuclear Research and Nuclear Energy (INRNE), Bulgarian Academy of Sciences. He was senior research scientist at INRNE and assistant professor at the Technical University of Sofia, Bulgaria. Later, he was a visiting Fulbright scholar at the Nuclear Engineering Department, Pennsylvania State University (PSU) and visiting scientist at the Research Center Rossendorf Inc., Germany. Prior to joining the NCSU faculty, he was a distinguished professor of nuclear engineering and graduate coordinator of nuclear engineering program at PSU. Dr. Kostadin Ivanov has been elected in February 2018 as a Chair of Working Party on Scientific Issues of Reactor Systems (WPRS) at Nuclear Science Committee (NSC) at NEA/OECD. WPRS leads the activities of four expert groups in reactor physics, fuel performance, multi-physics, fuel cycles, verification and validation, and uncertainty analysis in modelling. Dr. Ivanov also actively participates and contributes to the activities of the Expert Group on Multi-Physics Experimentation, Benchmarking and Validation (EGMPEBV) at NSC. For about twenty years Dr. Ivanov has led and continues to lead international reactor physics and multi-physics benchmark activities at NSC/NEA.
Professor, Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), South Korea.
Challenges for Autonomous Water-cooled Small Modular Reactor ATOM Development
This presentation is about R&D activities for an advanced autonomous PWR-type SMR (ATOM) under development at CASMRR (Center for Autonomous SMR Research) at KAIST. Beginning from motivations and fundamental characteristics of the autonomous nuclear reactors, a PWR-type SMR will be discussed in view of the reactor core and system designs for autonomous operations. To achieve a high level of safety and autonomous operation, the soluble boron should be removed from PWR and passive load-follow capability will be dramatically improved by a soluble-boron-free (SBF) SMR design. Design features and performances of the SBF ATOM core designs will be discussed in detail. Development and applications of the AI (Artificial Intelligence) to the autonomous SMR will be also introduced and discussed. For the ATOM system, an AI system called GAIA is being developed for multi-purpose applications including control, monitoring and diagnosing etc. Lastly, the major issues and challenges for the autonomous SMR development will be addressed as well.
Dr. Yonghee Kim is currently an associate professor at Korea Advanced Institute of Science and Technology (KAIST) in Korea. He has a B.Sc in Nuclear Engineering from Seoul National University and received his M.Sc and Ph.D in Nuclear Engineering from KAIST in 1990 and 1995, respectively. His research interests include advanced reactor concepts and development of nuclear reactor analysis methods. One of his major researches is the transmutation of nuclear wastes such as TRUs (Transuranics) and long-lived fission products in advanced nuclear reactor systems. He is also working on a new innovative photo-transmutation of nucleus by using the LCS (laser Compton scattering) photons. He was a senior researcher at KEPRI (Korea Electric Power Research Institute), principal researcher at KAERI (Korea Atomic Energy Research Institute), visiting scientist at Argonne National Laboratory and associate professor at Ulsan National Institute of Science and Technology, before joining KAIST. He has extensive research and management experiences in the area of advanced reactors and reactor physics. He is a member of KNS (Korean Nuclear Society) and ANS (American Nuclear Society). From 2013 to 2015, he served as a member of advisory committee for NSSC (national safety and security commission) of Korea. He is an editorial member of NET (Nuclear Engineering and Technology, official journal of KNS. He is also director of a government-funded research center for autonomous SMR research (CASMRR) at KAIST. From 2017, he is the chair of the Reactor Physics Division of KNS.
Professor Emeritus, Royal Military College (RMC), Canada.
Modelling and Simulation of Radiation Exposure in Space
Dr. Brent Lewis graduated from the University of Toronto with a BSc degree in Physics, MEng in Aerospace Science and Engineering and PhD in Nuclear Engineering. Dr. Lewis is currently an Emeritus Professor at the Royal Military College of Canada (RMC) and a Senior Consultant for the nuclear and space industry. He retired in 2015 as the Dean of the Faculty of Energy Systems and Nuclear Science at the University of Ontario Institute of Technology and was previously Professor of Nuclear Engineering and an NSERC/UNENE/COG Industrial Research Chair at the RMC. Prior to his academic appointments he was a Research Scientist and Section Head of Fuel Modelling at the Chalk River Laboratories of Atomic Energy of Canada Limited. Over the past 37 years his research has principally focused on nuclear fuel behaviour and aircrew/spacecrew radiation exposure, where he has published over 360 papers and reports, including a comprehensive textbook for Wiley on the “Fundamentals of Nuclear Engineering” in 2017. He has co-supervised/supervised 25 MASc and 22 PhD students. He received several literary awards from the American Nuclear Society and the Research Excellence Prize at RMC. He is also a fellow of the Canadian Nuclear Society and received the 2013 Outstanding Contribution award from the Canadian Nuclear Society.
Director of Nuclear Science & Technology’s Modeling and Simulation, Idaho National Lab (INL), United States.
Dr. Richard Martineau began employment at the Idaho National Laboratory (INL) in July of 1989. Dr. Martineau has twenty-nine years’ experience conducting numerical methods R&D and computational engineering investigations. Expertise includes computational fluid dynamics, nonlinear coupling methods for multiphysics applications, compressible material dynamics, fluid dynamics and heat transfer theory, and thermodynamics.
In 2008, Dr. Martineau’s Laboratory Directed Research and Development (LDRD) project enabled the development of INL’s Multiphysics Object-Oriented Simulation Environment (MOOSE) computational framework. The development of MOOSE resulted in a multitude of MOOSE-based code efforts funded by Department of Energy (DOE). MOOSE-based applications include multi-scale materials, reactor physics and radiation transport, thermal fluids, advanced manufacturing, subsurface chemistry, geophysics, among others. The MOOSE framework was given an LGPL Version 2.1 Open Source Software in March of 2014 and received R&D Magazine's 2014 R&D 100 Award. The framework has been downloaded (https://github.com/idaholab/moose) approximately 25,000 times and MOOSE-based applications are being developed around the world. Dr. Martineau is the primary developer of the PCICE (Pressure-Corrected Implicit Continuous-fluid Eulerian) time integration method for compressible flows and is leading the effort to develop the MOOSE-based application called Bighorn, which is designed to simulate multidimensional single- and two-phase conjugate heat transfer domains. He is also the principle investigator for RELAP-7 (also MOOSE-based), the next generation nuclear systems analysis/safety code.
Dr. Martineau is currently the Director of Nuclear Science & Technology’s (NS&T) Modeling and Simulation and is responsible for those aspects of developing advanced numerical methods, scientific numerical packages, high-performance computing frameworks, and multiphysics analysis tools for nuclear power applications at the INL. Under Dr. Martineau’s leadership, DOE’s Office of Nuclear Energy (DOE-NE) funded modeling and simulation efforts in INL’s NS&T Directorate have progressed from zero funds in FY-2008 to more than sixteen million dollars in FY-2016.
Professor, NSERC-UNENE Industrial Research Chair
Issues and Experiences in Multi-scale Uncertainty Propagation in Computational Best Estimate Plus Uncertainty Approaches
State-of-the-art safety analysis methods employ best-estimate computer simulations and include estimations of the effect of uncertainties. Such analyses may demonstrate improved margins as compared to traditional deterministic methods and provide a more realistic understanding of event propagation. The major elements required in the analysis are i) suitable inlet and boundary conditions, ii) computer codes which have various internal models and modelling parameters, iii) input files describing the plant geometry and conditions, and iv) uncertainties in the models, inputs and boundary conditions. In general the fidelity of these simulations and uncertainties has been improving since the first best estimate treatments in the 1980s and 1990s. International activities such as the recent OECD-NEA benchmark activities and BEMUSE project have examined various uncertainties and their treatment. While many papers exist on the results or methods used to propagate uncertainty, few focus on the fidelity of the input uncertainties, namely item (iv) above. In particular the uncertainty results may be dependent on the number of variables involved, the degree which co-variances are known or can be established, and the fidelity of the uncertainties assigned to each variable. An emerging topic of discussion is the interdependency of some uncertainties that are assigned for microscopic level transport phenomena (such as interfacial shear or heat transfer) and more macroscopic, or combined effect variables like Critical Heat Flux and Post Dryout heat transfer. This presentation specifically examines some of the key elements of these projects with specific emphasis on the role of microscopic and macroscopic input uncertainty quantification.
Director of Nuclear Power Software Development Center ofCentral Research Institute, State Power Investment Corporation (SPIC), China.
The Progress of COSINE Development and Assessment
To promote NPP software self-reliance development, DOE of China approved NPP software R&D key project named as COSINE. The scope of this project includes reactor physics, thermal hydraulics, safety analysis, severe accidents, probability safety assessment and etc. At this stage, the lattice physics code (cosLATC), core analysis code (cosCORE), neutron kinetics code (cosKIND), subchannel analysis code (cosSUBC), system safety analysis code (cosSYST) and containment safety analysis code (cosCONT) are under development in SNPSDC. COSINE is an integrated platform and modular code system for NPP design and analysis. Multi-physics, multi-dimensional code coupling, user-friendly GUIs, sensitivity/uncertainty analysis, parallel computation scheme will be introduced in COSINE. In the later stage, more advanced modules, functions and technologies are expected to be easily added to the code package. A brief introduction of COSINE code package, include the code architecture, the theoretical physics model and the current progress, and the verification and validation results are presented. Currently, most of the key design work is accomplish. The validation results show that COSINE met well with the benchmarks and experiment data.
Feng SHEN, Dr. and Prof., Director of Nuclear Power Software Development Center of Central Research Institute, State Power Investment Corp. He received his Ph.D of nuclear energy science and technology and worked as operator of Heavy Water Research Reactor (HWRR) before 2000. From 2000, he worked for the research and design of the China Advanced Research Reactor (CARR). From 2010, he focused on the Chinese self-reliance development of nuclear power plant design software-COSINE, and focused on the research of advanced nuclear energy system and safety research, such as small modular integrated reactor, fourth generation nuclear energy system and so on.
High Performance Computing at CNL
Dr. Radford has a B.Sc. in Applied Mathematics and Physics, a M.Sc. in Mechanical Engineering, and a Ph.D. in Mechanical Engineering. His research for his Ph.D. was performed at AECL on the high strain rate properties of CANDU pressure tube material. Darren is currently the Director of the Fuels, Materials & Design Division at Canadian Nuclear Laboratories (formerly Atomic Energy of Canada Ltd.- AECL), Chalk River, Ontario. In this role, he has oversight of five Departments performing computational and experimental research in areas including nuclear fuels, mechanical equipment design, plant monitoring and dynamics, fracture mechanics, and materials sciences. Darren started working in the Canadian nuclear industry 30 years ago at Whiteshell Laboratories, and moved to Chalk River Laboratories in 2005. From 2000 to 2005 Dr. Radford worked at Cambridge University doing research and teaching at the Cavendish Laboratory, and the Engineering Department (in the Mechanics, Materials and Design Division). While in Cambridge, Darren was a Fellow of Pembroke College and performed computational and experimental research on the high strain rate behaviour of materials and structures.
SCALE Deputy Manager, Multi-Physics Team Lead, Oak Ridge National Laboratory (ORNL), United States.
Interfacing SCALE with the Future
In this talk, we will begin by sharing ideas on what constitutes a good simulation interface, discussing both interfaces for humans and interfaces for other computer programs. Humans need concise interfaces that allow them to see the big picture, especially when systems and models get complex. Computer programs need rigid data structures that can easily be checked for validity and unambiguously parsed and processed. Although there is probably no single interface that is optimal for both humans and computers, we propose declaring a target for a particular interface is essential. Practical examples from SCALE will be used as the backdrop, including capabilities for "conventional" thermal-spectrum reactors (e.g. CANDUs) and new, "advanced" reactors (e.g. molten salt reactors, sodium-cooled fast reactors, and gas-cooled pebble bed reactors). In closing, we will highlight capabilities and interfaces slated for the next release of SCALE, to be used at U.S. Nuclear Regulatory Commission for advanced reactor confirmatory calculations.
Dr. William Wieselquist is a senior research and development staff member and leader of the multi-physics team within the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory (ORNL). Since joining ORNL in 2012, he has become heavily involved in the SCALE code system, now holding roles of SCALE deputy manager, lead developer of the ORIGEN depletion/decay code, lead developer of the Sampler uncertainty quantification code, and co-developer of the Polaris lattice physics code. Dr. Wieselquist is also an active member of the SCALE training team, designing and leading training at ORNL and through the OECD/NEA. Outside of SCALE, he participates in DOE CASL, NEAMS, and Nuclear Data Benchmarking programs and maintains involvement in OECD/NEA international benchmarks.
Dr. Wieselquist received a Ph.D. in Nuclear Engineering from North Carolina State University in 2009. From 2009 to 2012 he worked at the Paul Scherrer Institut, where he established an uncertainty quantification platform around the Studsvik Scandpower core analysis codes in addition to core follow and safety analysis duties for the Swiss nuclear power plant, Beznau Units 1 and 2.
Dr. Wieselquist’s main research interests include sensitivity analysis, uncertainty quantification, data assimilation, verification and validation, lattice physics, depletion methods, multi-physics coupling, as well as user and application interface design.
Professor, University of Saskatchewan, Canada
Accident Tolerant Fuel and Novel Cladding to Address Development of GenIV and Small Modular Reactors
J.A. Szpunar joined the Department of Mechanical Engineering at the University of Saskatchewan in August 2009, as Tier I Canada Research Chair. He came from McGill University where he was Professor of Materials Science and Birks Chair in Metallurgy. His research interests extend to various areas of materials related investigations. More recently, his research has focused on environmentally friendly energy generation, in particular the extraction and purification of hydrogen, accident tolerant nuclear fuel and research on advanced materials for Generation IV nuclear reactors. His research supports also various clean energy programs, research on more safe and secure materials for oil and gas transportation.
Dr Szpunar has a strong record of research productivity. 42 PhD students and 27 MSc. students graduated under his supervision. He is an author and co-author of more than 900 research papers.
Development and use of accident tolerant fuel (ATF) in commercial light water reactors (LWRs), Generation IV nuclear reactors and small modular reactors (SMR) is studied extensively at present. This presentation will address research of our team in area of new high thermal conductivity composites based on urania, thoria and silicates and diborrite of uranium and thorium. Presented work is both experimental and theoretical.
We manufactured various (collaboration with UBC) types of composites with uranium and thorium, in also uranium carbides, silicates and borate and often obtained very significant improvement of thermal conductivity. Performed for the first time detailed microstructural analysis and comparative studies of influence of porosity and fission products on the thermal conductivity.
Our computational prediction were based on Density Functional Theory (DFT) and were focused on prediction of thermal conductivity, electronic properties and structural changes of materials for fuels at very high temperatures, and under irradiation and also in oxidation environment. DFT and molecular dynamics (MD) was used to predict the thermal conductivity and mechanical properties of defects and incorporation of xenon (Xe) and zirconium (Zr) fission products in studied in materials for fuel. Materials like UO2-BeO, UO2-SiC and uranium carbide, uranium silicates, uranium boride were studied. This work provides fundamental understanding of structure-property relationships under irradiation and with presence of fission products in fuels and can serve as an important data for future experimental efforts. These calculations were performed within the framework of (DFT) +U approach, using Quantum ESPRESSO (QE) code.
Our research is also focus on fuel cladding and address the problem of severe environment of service in Gen IV nuclear reactors, where high temperature, supercritical water oxidation and other factors may require the replacement of Zr alloys in many applications. We propose novel coatings to improve performance of Zr alloys and also investigated stainless steels, nickel alloys and Inconels
as a possible replacement of Zr alloys. Microstructural design of materials for fuel cladding allowed us to improve very extensively resistance to supercritical water oxidation. GenIV and SMR reactors will require special materials that have to serve in a very critical environment and should improve the safety and reliability of future reactors.