Advances in Nuclear Fuel Management V
Advances in Nuclear Fuel Management V March 29 - April 1, 2015 Hilton Head Island, SC Official Program http://anfm2015.org ANFM 2015 - Advances in Nuclear Fuel Management V March 29 - April 1, 2015 Foreword Welcome to the American Nuclear Society’s Advances in Nuclear Fuel Management topical meeting. The 2015 edition of this meeting is the fifth in the series, held every six years. It has become a tradition to hold the meeting in early spring in sunny Hilton Head, South Carolina. We hope you enjoy everything the area has to offer. Please let any member of the organizing committee know if you have any questions or concerns. Volunteers from the Columbia, SC Local Section will be wearing navy blue polo shirts, and they will be happy to help you with information about the local area. Thank you for attending and supporting the important work of the American Nuclear Society and all our co-sponsoring international organizations. Bill Herwig, General Chairman Damon Bryson, Assistant General Chair Acknowledgement The organizing committee would like to thank the following: • • • • • • • • Our financial sponsors for their generous contributions to the success of the meeting. Please take the time to thank them for their commitment. Our co-sponsoring technical societies and ANS technical divisions for helping us to publicize the meeting. The technical program committee for soliciting papers, organizing and performing paper reviews, and chairing our technical sessions. Ms. Hanna Shapira (TICS) for management of the website, registration, program booklet, and CD. Our panel and plenary session speakers, who were willing to accept another assignment on top of an already hectic schedule. RSICC - Radiation Safety Information Computational Center (ORNL) for the duplication of the CDs. The Consortium for Advanced Simulation of LWRs (CASL) for co-locating their workshop at the end of the conference. Finally, I would personally like to thank the organizing committee, who have done an outstanding job of coordinating the program and the meeting arrangements. What we lacked in face-to-face meetings we made up for in e-mails, faxes, conference calls, and every other form of communication possible to get the meeting planned. The hours you spent ensuring a successful meeting are very much appreciated; the meeting simply wouldn’t have happened without your dedication. Bill Herwig, General Chairman Damon Bryson, Assistant General Chair 1 ANFM 2015 - Advances in Nuclear Fuel Management V March 29 - April 1, 2015 Organizing Committee General Chair Assistant General Chair Past Chair & Advisor Technical Program Co-chair Co-chair Co-chair Publications Finance Registration/Web Apps Publicity Sponsors/Fundraising Tours/Events Banquet Coordinator Workshops Student Volunteer Coordinator Bill Herwig, SCE&G Damon Bryson, SCE&G John Siphers, Progress Energy Ivan Maldonado, University of Tennessee Atul Karve, Global Nuclear Fuel Ron Ellis, ORNL Elise Malek, Westinghouse Duane Twining, SCE&G Hanna Shapira, TICSs Lisa Marshall, NC State University Sarah Gillham, Southern Nuclear Caroline Duncan, Westinghouse Courtney Tampas, SCE&G Rose Montgomery, TVA Matthew Presson, SCE&G Jamel Bell, SCE&G Bill Herwig Damon Bryson John Siphers Ivan Maldonado Atul Karve Ron Ellis Elise Malek Duane Twining Hanna Shapira Lisa Marshall Sarah Gillham Caroline Duncan Courtney Tampas Rose Montgomery Matthew Presson Jamel Bell 2 ANFM 2015 - Advances in Nuclear Fuel Management V March 29 - April 1, 2015 Technical Program Committee Tunc Aldemir, OSU, USA Mehdi Asgari, Studsvik Scandpower, USA Brian Aviles, KAPL, USA Paul Bailey, Duke, USA Steve Baker, Transware, USA Jeff Borkowski, Studsvik Scandpower, USA Robb Borland, First Energy, USA Steve Bowman, ORNL, USA Jeff Bradfute, Westinghouse, USA Juan Casal, Westinghouse, Sweden Angelo Chopelas, GNF, USA Dimitrios Cokinos, BNL, USA Chris Comfort, Southern, USA Edmundo Del Valle Gallegos, IPN, Mexico Arthur DiGiovine, Studsvik Scandpower, USA Alex Dolgov, TVEL, Russia Tom Downar, Univerity of Michigan, USA Mike Dunn, ORNL, USA Troy Eckleberry, TVA, USA Bob Einziger, NRC/US, USA Ronald Ellis, ORNL, USA Filip Fejt, Technical Univ at Prague, Czech Rep Fausto Franceschini, Westinghouse, USA Juan Luis Francois, UNAM, Mexico Norman Garner, AREVA, USA Kenneth Geelhood, PNNL, USA Jess Gehin, ORNL, USA Ali Haghighat, UFL, USA Ayman Hawari, NCSU, USA Charles Heck, GNF, USA William Herwig, SCANA, USA Jim Hoerner, AREVA, USA Nadine Hollasky, Bel V, Belgium Clive Ingram, Office of Nuclear Regulation, United Kingdom Victor Inozemtsev, IAEA, Int. Kostadin Ivanov, PSU, USA John Jones, Office of Nuclear Regulation, United Kingdom Atul Karve, GNF, USA Doddy Kastanya, Candu Energy, Canada Paul Keller, Areva, USA Hany Khalik, Purdue University, USA Travis Knight, USC, USA Dave Knott, Studsvik Scandpower, USA Tomasz Kozlowski, University of Illinois, USA Dave Kropaczek, Studsvik Scandpower, USA Vefa Kucukboyaci, Westinghouse, Maria Teresa Lopez Carbonell, IBERDROLA, Spain Ivan Maldonado, University of Tennessee, USA Christian Malm, Vattenfall, Sweden Cecilia Martin-del-Campo, UNAM, Mexico K Matsuura, NFI, Japan Vaclav Mecir, CEZ, CZECH Ugur Mertyurek, ORNL, USA Mitch Meyer, INL, USA Pierre Mollard, AREVA, France Rob Montgomery, PNNL, USA Rose Montgomery, TVA, USA Brian Moore, GNF, USA Bruce Morgen, Duke, USA Rahim Nabbi, Juelich Research Center, Germany Eleodor (Dorin) Nichita, UOIT/AECL, Canada Chvala Ondrej, University of Tennessee, USA Shinji Ono, Westinghouse/NFI, Japan Abderrafi Ougouag, INEL, USA Mohamed Ouisloumen, Westinghouse, USA Scott Palmtag, Core Physics, USA Bojan Petrovic, GA Tech, USA Dubravko Pevec, University of Zagreb, Croatia Trent Primm, Primm Consulting, USA Zhao Qiang, Harbin Eng University, China Manuel Quecedo, ENUSA, Spain Farzad Rahnema, GA Tech, USA Sumit Ray, Westinghouse, USA Tony Reese, GNF, USA Michael Reitmeyer, Exelon, USA Javier Riverola, ENUSA, Spain Kan Sakamoto, NFD, Japan Alain Santamaria, CEA, France Hitoshi Sato, GNF, Japan Jeff Secker, Westinghouse, USA Koroush Shirvan, MIT, USA John Siphers, Duke, USA Steve Skutnik, University of Tennessee, USA Russell Stachowski, GNF, USA Bob StClair, Duke Energy, USA Marco Streit, Paul Scherrer Institut, Switzerland John Strumpell, AREVA, USA Scott Thomas, Duke Energy, USA Jim Tulenko, UFL/US, USA Paul Turinsky, NCSU, USA Michael Tusar, Exelon, USA Tadashi Ushio, NFI, Japan Mojmir Valach, NRI, Czech Republic Nicolas Waeckel, EDF, France Fu Xiangang, CGNPC, China Peng Xu, Westinghouse, USA Akio Yamamoto, Nagoya University, Japan Masatoshi Yamasaki, NFI, Japan Koo Yang-Hyun, KAERI, Korea Ying Yi, SNERDI, China Serkan Yilmaz, GNF, USA Hiroyuki Yoshida, Toshiba, Japan Quun Zee, KAERI, South Korea Hongbin Zhang, INL, USA Jinzhao Zhang, Tractebel Eng / GDF SUEZ, Belgium A special thank you and recognition for the following individuals for being engaged, flexible, and for volunteering the time in completing the many comprehensive reviews the Technical Program Co-chairs solicited of them. Their constructive feedback review ensured that the papers continue to be of exceptional quality and live up to the high standards for this conference. Christian Malm, Vattenfall Norman Garner, Areva Jim Hoerner, Areva Vefa Kucukboyaci, Westinghouse John Jones, Office of Nuclear Regulation, UK Doddy Kastanya, Candu Energy, Canada Tomasz Kozlowski, University of Illinois, USA Russell Stachowski, GNF, USA Koroush Shirvan, MIT, USA 3 Dimitrios Cokinos, BNL, USA Masatoshi Yamasaki, NFI, Japan Jeff Secker, Westinghouse, USA Akio Yamamoto, Nagoya University, Japan Nadine Hollasky, Bel V, Belgium Advances in Nuclear Fuel Management V 29 - April 1, 2015 ANFM 2015 - Advances in Nuclear Fuel Management V March March 29 - April 1, 2015 Financial Sponsors HOME Call for Papers Program Login Registration Hotel Transportation Area Info. Committees HOME Call for Papers Program Login Registration Hotel Transportation Area Info. Committees Platinum Gold Bronze Other Sponsors • American Nuclear Society (ANS) • ANS - Eastern Carolinas Section (ECS) • ANS - Columbia SC Section • ANS - Reactor Physics Division • ANS - Fuel Cycle and Waste Management Division • ENS - European Nuclear Society • CNS - Canadian Nuclear Society • AESJ - Atomic Energy Society of Japan • KNS - Korean Nuclear Society • SNM - Mexican Nuclear Society • OECD/NEA - Organizatin for Economic Cooperation & Development/Nuclear Energy Agency 4 ANFM 2015 - Advances in Nuclear Fuel Management V March 29 - April 1, 2015 General Information Registration Session Chair Information Registration is required for all attendees and presenters. Badges are required for admission to all events. Please complete and return a “Session Chair Sign-in Form.” Please attend the Breakfast (7-8am) on the day of your session and be present at your session room at least 15 minutes prior to the start of the session. This will allow you to greet and coordinate media arrangements with the speakers, as well as collect biographical sketches. For the sake of meeting attendees, PLEASE keep the session perfectly synchronized as shown in this final program. For “no shows” simply adjourn the session at the next allotted time (i.e., don’t shift papers to earlier slots to fill a void). You may find it helpful to bring your own laptop and upload the speakers’ presentations during the breakfast or pre-session meetings. Alternatively, please ensure there is a laptop available for facilitating presentations during your entire session. You will have a student assistant to assist you during the meeting. You may use his/her assistance to drive the presentation, help with A/V etc. He/She will checkin with you prior to the start of the session and ask you to sign a confirmation of assistance at the end of the session. The Full Conference Registration Fee for Member and Non-Member includes: All technical sessions, CD of all proceedings, coffee breaks/snacks, and Sunday night reception. Registration does not include Monday night dinner or Tuesday night banquet. The One Day Conference fee for Member and NonMember includes: All technical sessions for the registration day, coffee breaks/snacks, and CD of proceedings. Does not include dinner ticket. The Student Registration Fee includes: All technical sessions, CD of all proceedings, coffee breaks/snacks, and Sunday night reception. Registration does not include Monday night dinner or Tuesday night banquet. Spouse/Guest includes: Sunday night reception, but does not include Monday night dinner or Tuesday night banquet. The Meeting Registration Desk is at Palmetto Landing: Sunday Monday-Wednesday 3:00 PM – 5:00 PM 7:15 AM – 8:00 AM Speaker Information Please sign the “Speaker Sign-in Form” at the registration desk. Note that the total time available for oral presentations (other than plenary and special sessions) is 25 minutes, so please check carefully the program for the time allocated to your presentation. We recommend that you allow for 1 to 3 minutes for questions and discussion at the end of your talk. Be alert, responsive, and respectful to other speakers when the Session Chair signals you that you’ve got 5 and 2 minutes remaining in your time slot. As a presenter you will be able to use your own laptop or bring a memory stick. It is your responsibility to check your presentation file for compatibility with the Session Chair. We recommend that you seek out and meet your Session Chair during the Conference Breakfast (7-8am each morning), and to also meet with the Session Chair at the presentation room 15 minutes before the session begins. Please provide a brief biography (name, organization, 2-3 line description of current work assignment) to the session chair prior to start of the session. Please check the signs and handouts for further information. Please contact any of the meeting organizers if you need help or have questions. You will have a student assistant to assist you during the meeting. You may use his/her assistance to drive the presentation, help with A/V etc. Please check with the session chair and the student assistant prior to the meeting. 5 ANFM 2015 - Advances in Nuclear Fuel Management V March 29 - April 1, 2015 Meeting Rooms 6 ANFM 2015 - Advances in Nuclear Fuel Management V Monday March 30, 2015 Plenary Session Palmetto Ballrooms A & B 8:00 AM Pierre Paul Oneid Senior Vice President & Chief Nuclear Officer Mr. Pierre Paul Oneid is Senior Vice President and Chief Nuclear Officer of Holtec International and the President of Holtec International subsidiary SMR, LLC. Mr. Oneid earned an Executive Master of Business Administration from Queens University in Canada in 1998 and a B.S. in Mechanical Engineering from the University of Ottawa, Canada in 1981. Dr. Russell Stachowski Chief Consulting Engineer - Reactor and Nuclear Physics Global Nuclear Fuel / GE Hitachi Nuclear Energy Dr. Stachowski is Chief Consulting Engineer-Reactor and Nuclear Physics for GE Hitachi Nuclear Energy. He has spent much of his career in the development, testing, licensing, and application of BWR fuel in the US, Mexico, Japan and Europe. He has led the Nuclear Methods team in developing lattice and core physics methods for BWRs. Russell holds a B.S. in Mechanical Engineering from the University of Notre Dame and M.S. and Ph.D. degrees in Nuclear Engineering from the University of California, Berkeley. Ronald Jones Vice President, New Nuclear Operations SCANA/SCE&G Has primary responsibility to identify, evaluate and structure investments in the development of two nuclear power plants from construction phase through commercial operation including completion of design, construction, start-up, testing, design verification tests, inspections, and transitioning to an operating organization. Responsibilities include satisfying all Nuclear Regulatory Commission and state and/or local governments’ requirements for the construction and operation of new nuclear power plants, budgeting, financing and recruitment. Identifies, evaluates and hires outside contractors as needed throughout all phases of development. Brian R. Beebe Director, Core Engineering, Engineering Center of Excellence, Westinghouse Electric Co. Brian R. Beebe is the Director of Core Engineering in Westinghouse Electric Company’s Engineering Center of Excellence. Westinghouse is the recognized world leader in the building of Nuclear Power Electric Generating Plants, Operational Support for Nuclear Power Plants, Nuclear Fuel Development and Supply and overall nuclear power generation research and development. Core Engineering has the responsibility for the design and operational support of PWR and BWR reactors across the globe. Brian is a three time recipient of the George Westinghouse Engineering Signature Award of Excellence, a five time recipient of the Performance Excellence Award, and a graduate of the Westinghouse Customer First leadership Program. During his tenure at Westinghouse Brian has worked at many of Westinghouse’s facilities worldwide including living for 2 years in Västerås, Sweden. Prior to joining Westinghouse Brian received his MS and BS with High Honors in Nuclear Engineering from the University of Florida. 7 ANFM2015 - Advances in Nuclear Fuel Management V Monday, March 30, 2015 Meeting Room: Sabal MM1 New Modeling Concepts, Reactivity Control, Generation of Cross Section Libraries and Whole Core Transport Calculations Chair: Hany Abdel-Khalik 10:00 60 Fuel Cycle Performance of Intermediate Spectrum Reactors with U/Th Feed and Continuous Recycling of U/TRU and Th/U3 Nicholas R. Brown Michael Todosow Brookhaven National Laboratory, Upton NY This paper documents fuel cycle analysis of an intermediate spectrum critical reactor with natural uranium and thorium as feed materials and continuous recycling of uranium with transuranics and thorium with uranium (mainly 233U). The objective of the effort was to determine the impacts of both intermediate spectrum and U/Th feed versus the reference fast spectrum and U-only feed case. In this context the intermediate energy regime is considered to be between 1 eV and 0.1 MeV. The potential benefits of introducing thorium into the natural resource feed include extending the availability of uranium resources as well as potential operational/safety benefits, particularly related to the void coefficient in some reactor designs. Two systems were analyzed at a near-equilibrium condition, a high void boiling water reactor with 54% of fissions occurring in the intermediate energy regime and a D2O cooled pressurized water reactor with 65% of fissions occurring in the intermediate energy regime. The systems analyzed in this study were determined to be self-sustaining at equilibrium when accounting for loss rates in fabrication and separations, and exhibit similar resource utilization when compared to a fast system. Differences in performance versus a fast system are driven to some extent by the fact that eta for 233U and 239Pu at intermediate energies is never as high as for 239Pu at fast energies. 10:25 75 Development of a Full-Core Reactivity Equivalence for FeCrAl Enhanced Accident Tolerant Fuel in BWRs Nathan M. George (1), Jeffrey J. Powers (2), G. Ivan Maldonado (1), Andrew Worrall (2), Kurt A. Terrani (3) 1) Department of Nuclear Engineering, University of Tennessee Knoxville, TN, 2) Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, Oak Ridge, TN, 3) Fusion and Materials for Nuclear Systems Division, Oak Ridge National Laboratory The impact of replacing Zircaloy with a candidate iron-chromium-aluminum (FeCrAl) -enhanced accident-tolerant fuel (ATF) cladding material was evaluated for modern 10 × 10 boiling water reactor (BWR) fuel bundles. The primary objective was to establish fuel design parameters for FeCrAl that match the reactivity lifetime requirements of standard Zircaloy bundles. To compare the neutronic effects of these fuel alterations against standard UO2/Zircaloy fuel, a method based on 2D lattice physics was established to estimate excess reactivity at the completion of each reactor operating cycle using a weighted fuel batch scheme. The methodology allows rapid scoping studies to be performed prior to full-core simulations. An additional paper [1] presents the lattice physics and 3D full-core results verifying the reactivity equivalence method for the alternate cladding fuel bundles. 10:50 89 Demonstration of a Full-Core Reactivity Equivalence for FeCrAl Enhanced Accident Tolerant Fuel in BWRs Nathan M. George (1), Jeffrey J. Powers (2), G. Ivan Maldonado (1), Andrew Worrall (2), Kurt A. Terrani (3) 1) Department of Nuclear Engineering, University of Tennessee, Knoxville, TN, 2) Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, Oak Ridge, TN, 3) Fusion and Materials for Nuclear Systems Division, Oak Ridge National Laboratory, Oak Ridge, TN The impact of replacing Zircaloy with a candidate iron-chromium-aluminum (FeCrAl)-enhanced accident-tolerant fuel (ATF) cladding material was evaluated for modern 10 × 10 boiling water reactor (BWR) fuel bundles. Lattice physics calculations were completed with the 2D deterministic codes SCALE/TRITON and CASMO, and 3D full-core calculations were performed with the NESTLE nodal diffusion code. The primary objective was to establish fuel design parameters for FeCrAl that match the reactivity lifetime requirements of standard Zircaloy bundles. Due to the high neutron absorption of FeCrAl relative to standard Zircaloy, the FeCrAl cladding and channel box thicknesses were decreased and the enrichment of uranium dioxide (UO2) fuel was increased. To compare the neutronic effects of these fuel alterations against standard UO2/Zircaloy fuel, a method based on 2D lattice physics was established to estimate excess reactivity at the completion of each reactor operating cycle using a weighted fuel batch scheme. This study is supported by a companion paper that thoroughly describes the established methodology [1]. With the cladding and channel box thicknesses halved, it was estimated that an average enrichment increase of 0.6% 235U throughout the fuel lattice would be required. Verification of this 2D reactivity method was performed with a 3D full-core parametric study. Matching the base UO2/Zircaloy cycle length of 527 effective full power days (EFPD) with UO2/FeCrAl required nearly the same fuel design adjustments in full-core calculations as were predicted by the lattice physics results, thus demonstrating the accuracy of the reactivity method. 11:15 41 BWR Control Rod Mechanical Design Considerations Based on a Review of General Electric Control Rod Design and Performance History Scott Nelson GE Hitachi Nuclear Energy, Wilmington, NC A review of GE/GEH BWR control rod design and performance history is conducted from the original equipment to the present time. This review identifies common causes of mechanical failures that are primarily attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). Identification of common causes allows for the establishment of several design considerations that should be incorporated into future BWR control rod designs, for the purpose of reducing the likelihood of cracking. Finally, the GEH UltraTM control rod design is reviewed, and the method by which each of the identified design considerations is incorporated into the new design is detailed. 8 ANFM2015 - Advances in Nuclear Fuel Management V Monday, March 30, 2015 11:40 81 Effect of Energy Group Structure on a Stylized European Pressurized Reactor (EPR) For Criticality Analysis Daniel Lago and Farzad Rahnema Nuclear & Radiological Engineering/Medical Physics Programs, Georgia Institute of Technology, Atlanta, GA, USA This paper investigates the effect of energy group structure on a stylized MCNP model of the European Pressurized Reactor (EPR). A benchmark of the EPR was previously developed purely for validation of transport methods. This study evaluates the cross sections generated by the lattice depletion code HELIOS for the EPR benchmark. This paper describes the generation of problem-specific multi-group cross sections in 2-, 4-, 8-, and 47- group structures with HELIOS, as well as initial results from assembly level MCNP calculations to evaluate the effect of group-collapsing. The paper also discusses the possible propagation of errors from the cross sections in whole-core calculations. Meeting Room: May MM2 Innovative Core Loading, Reload Design, and Licensing Chair: Rodolfo Ferrer 10:00 7 Experience Developing Power Peaking Penalties for Fuel Assemblies Reconstituted with Stainless Steel Rods at Oconee Nuclear Station David Orr and Joy Forster Duke , Charlotte, NC Fuel assembly to core baffle interactions are a known phenomenon in the utility and vendor community. Currently at the Oconee Nuclear Station, which is owned and operated by Duke Energy Carolinas, there are two primary fuel mechanical concerns involving fuel assembly to core baffle interaction: spacer grid wear and fuel rod wear. While the mechanism of wear is different between the two issues, both may lead to concerns about the integrity of the fuel if no mitigating actions are taken. Hence, Duke Energy has chosen to perform fuel assembly reconstitution—the insertion of stainless steel rods into fuel assemblies that already have resided in the core for at least one cycle of operation—in a proactive fashion to mitigate the risk of fuel failures associated with these mechanisms of wear. The decision to reconstitute assemblies raises the question of how to address the impact to power peaking and, subsequently, the validity of the safety analyses and maneuvering analysis for a given cycle with potentially 50 -60 stainless steel rods present in the core. One way is to create rod peaking penalties by modeling the affected fuel assemblies, including their burnup history, both with and without the insertion of the stainless steel rods. A comparison is made between the predicted rod power peaking in both cases, and judgments may be made about appropriate penalties to be applied in subsequent analyses. 10:25 8 Innovative Approach to Reloading an Initial Cycle Jun Shi ,Samuel Levine, and Kostadin Ivanov The Pennsylvania State University (PSU), University Park, PA The objective of this paper is to present analyses of an innovative approach to reload an initial cycle loading pattern (LP) of a PWR by selecting the reload pattern fuel assemblies, FAs, based on their K∞ rather than on their initial enrichment. In this new method, the FA K∞ is the primary selection factor, i.e., the FAs having the lowest K∞ are discarded after each cycle. However, it has been discovered that the sum of the 235U and the 239Pu nuclide’s number densities are also very important factors when choosing the used FAs to be reloaded in cycle 2. The Haling Power Depletion (HPD) method has been extensively used to guide the design of the reload core. It was discovered that the first cycle has to be the highest possible leakage core because of the nature of the condition of the end-of-cycle FAs. The HPD acceptable reload design ended up with large loss in cycle length. In fact, the second cycle is also a relatively high leakage core. A more accurate step depletion calculation is implemented afterwards to verify the design. The simple relative fuel cost comparison made between the two methods for the first reload cycle must now be made for the second cycle. Studies will continue on reloading further cycles to obtain a long term understanding of minimizing the fuel cost. 10:50 14 On Multiobjective Optimisation Approaches for In-Core Fuel Management Optimisation Evert B. Schlünz (1, 2), Pavel M. Bokov (1), Jan H. van Vuuren (3) 1) Radiation and Reactor Theory, Necsa, Pretoria, South Africa, 2) Department of Logistics, Stellenbosch University, Matieland, South Africa, 3) Department of Industrial Engineering, Stellenbosch University, Matieland, South Africa In the in-core fuel management optimisation (ICFMO) problem, a fuel reload configuration is sought which optimises the performance of a nuclear reactor, while also satisfying prescribed operational constraints. ICFMO has been studied for several decades, initially as single-objective optimisation problems but in recent years also as multiobjective optimisation problems. Several of the multiobjective ICFMO approaches adopted in the literature, however, exhibit serious shortcomings or drawbacks, and very little research has been performed to address or overcome them. In this paper, we present a brief overview of the various multiobjective ICFMO approaches found in the literature. We also provide a commentary on what we believe to be the most important shortcomings and drawbacks in these approaches and, concurrently, present our suggestions for addressing them. The workability of these suggestions are demonstrated by their application to two test problems for the SAFARI-1 research reactor. The results indicate that our suggested approaches are indeed feasible for multiobjective ICFMO problems. The aim of this paper is to encourage further research toward multiobjective ICFMO via the inclusion of sound principles from the field of operations research. 9 ANFM2015 - Advances in Nuclear Fuel Management V Monday, March 30, 2015 11:15 17 The Greedy Exhaustive Dual Binary Swap Method for Fuel Loading Optimization Using the Poropy Reactor Optimization Tool Carl C. Haugen and Kord S. Smith Massachusetts Institute of Technology, Cambridge, MA This paper presents a deterministic optimization scheme termed Greedy Exhaustive Dual Binary Swap for the optimization of nuclear reactor core loading patterns. The goal of this optimization scheme is to emulate the approach taken by an engineer when manually optimizing a reactor core loading pattern. This is to determine if this approach is able to locate high quality patterns that, due to their location in the core loading solution space, are consistently missed by standard stochastic optimization methods such as those in the simulated annealing class. This optimization study is carried out using the poropy tool to handle the reactor physics model. Optimizations of the full depletion problem result in the deterministic Dual Binary Swap optimizer locating patterns that are of higher quality than those found by the stochastic Simulated Annealing optimizer, with comparable frequency. The Dual Binary Swap optimizer is, however, found to be very dependent on the starting core conguration, and can not reliably nd a high quality pattern from any given starting conguration. Meeting Room: Palmetto B MM3 PANEL Discussion - "Small Modular Reactors: Challenges and Proposed Solutions to Successful Deployment" Chair: Andrew Worrall 10:00 Objective of panel: To provide an insight from a number of perspectives as to the challenges facing the future deployment of Small Modular Reactors (SMRs), particularly of the integral PWR (iPWR) variety. These challenges include technical, regulatory, economic, and supply chain issues. The intent is that the speakers identify the challenges and propose solutions. 10 ANFM2015 - Advances in Nuclear Fuel Management V Monday, March 30, 2015 Meeting Room: Sabal ME1 Error Quantification of Core Simulation Capabilities, Core Follow Data to Enhance Core Simulation Fidelity, Utilization of Zero Power Physics Tests Chair: Fausto Franceschini 4:20 PM 83 Tutorial Series on Characterization of Uncertainty (TUSC): Reduced Order Modeling, Dimensionality Reduction, Surrogate Modeling, Function Approximation, Fitting, etc. Hany S. Abdel-Khalik School of Nuclear Engineering, Purdue University, West Lafayette, IN The increased reliance on modeling and simulation for the analysis of complex engineering systems has made it essential to devise scientifically defendable approaches for the characterization of uncertainties. The last two decades have witnessed the development of many reliable and efficient methods capable of identifying, quantifying, prioritizing, and ultimately reducing the various sources of uncertainties. Despite the recent theoretical triumphs, it is safe to say that these developments are not readily accessible by engineering practitioners who are the intended beneficiaries of these developments. This is because the subject of uncertainty characterization is heavily mathematical in nature, which requires rigorous and abstract mathematical constructs to properly describe it in its most general form, which is the preferred approach by the developers, mainly mathematicians and statisticians. While the rigor is definitely needed, it has made it extremely difficult for practitioners to understand the mechanics of the various methods and independently evaluate their advantages and limitations, instead of relying on expert-judgment. The TUSC series intends to break this pattern by introducing the material in a form more accessible by engineers and engineering practitioners. Our introduction will favor intuition over rigor, and will provide enough intuitive arguments, as supported by reasonable amount of rigor, to help reveal the core ideas behind each method. This will enable engineers further develop and customize the methods for their own needs. The present manuscript introduces the basic concepts and more importantly the distinguishing factors between reduced order modeling, dimensionality reduction techniques, surrogate modeling; all basic ingredients of uncertainty characterization methods. 4:45 PM 87 Identifying Modeling Parameters to Influence an Operating Experience Observation Atul A. Karve and Russell E. Stachowski Global Nuclear Fuel, Wilmington, NC Recent observations of a boiling water reactor reload cycle operation have reinforced the need for robust core simulator methods. Specifically, these methods can be challenged in predicting operating parameters that are monitored by adaptive methods (derived core observables). To manage an unanticipated behavior in these derived core observables, either excess conservatism in design needs to be incorporated and / or mitigating actions for adverse operation need to be exercised. Such actions are undesirable because the excess margin and / or the operating changes can adversely impact the overall fuel cycle economics. Therefore, there is ever more need for methods to be able to design the reload cycle (referred to as the offline prediction) such that when the reload cycle is operated closely as designed, the core monitoring system (referred to as the online prediction) should be consistent with the offline, i.e. the derived core observables do not significantly depart from the design. Anomalies occur when there is an unusual unexplainable deviation in the derived core observables. While it is reassuring that this is an isolated occurrence, the particular deviation becomes an operating experience observation that needs to be further analyzed and studied. This paper attempts to do that for one such deviation – the derived core observable relates to the maximum fraction of linear power density (MFLPD). In an INPO operating experience in 2014, the MFLPD was observed to significantly deviate between the online and offline predictions. This study addresses that deviation by identifying specific causal factors in the modeling that can be adjusted to obtain the observed effect. The purpose is to model the hypothesized behavior that the method can capture self-consistently. Such enhanced modification is not necessary to be generalized; however, it attempts to capture plant specific operating variations (that are unknown) and / or embody realistic phenomena (that are known but possibly not sufficiently modeled). In the end, this exercise identifies areas for further study to improve the prediction that could alleviate some of the operating uncertainty that needs to be incorporated as part of the design. 5:10 PM 40 Core Follow and Cold Critical Calculations of Operation Cycles After Extended Outage in BWRs Tsuyoshi Ama, Takashi Yoshii, Akihiro Fukao (1), Katsuyoshi Oyama (2) Nuclear Core Engineering Dept., TEPCO SYSTEMS CORPORATION (TEPSYS), Koto-ku, Tokyo, Japan, 2) Nuclear Power Plant Management Dept. Tokyo Electric Power Company (TEPCO), Chiyoda-ku, Tokyo, Japan Core follow and cold critical calculations of several cycles including cycles after extended outage are performed for Japanese commercial BWRs using CASMO-4/SIMULATE -3. The core follow calculation results and cold eigenvalues in cycle after extended outage are compared to those in cycles after normal outage. Hot eigenvalues and traversing in-core probe (TIP) route mean square (RMS) errors are evaluated in the core follow calculation. The comparison shows that the change of reactivity caused by extended outage is properly evaluated, and that the TIP RMS errors are similar between the cycles after normal outage and extended outage. The cold eigenvalues are also evaluated in the cold critical calculation. The levels of cold eigenvalue in cycle after extended outage are similar to those in cycles after normal outage. These comparison results show that CASMO-4/SIMULATE-3 is applicable to the evaluation in cycle after extended outage. 5:35 PM 74 Evaluation of the NPP Krško Core by JSI and Westinghouse Nuclear Analysis Codes Marjan Kromar (1), Fausto Franceschini (2), Dušan Ćalić (1), Harish C. Huria (2) 1) Jožef Stefan Institute, Reactor Physics Division, Ljubljana, Slovenia, 2) Westinghouse Electric Company LLC, Cranberry Township, PA Jožef Stefan Institute (JSI) and Westinghouse have performed reactor physics analysis of the several NPP Krško cycles using their respective core simulator packages, e.g. CORD-2 and NEXUS/ANC 9. This paper shows the performance of each core simulator to predict the plant physics behavior, specifically analyzing the comparison vs. measurements for the key core parameters. Critical boron concentrations, control rods worth and isothermal temperature coefficient are compared to the measured values. The results show satisfactory performance from both code systems and their adequacy to support the core design calculations and fuel loading optimization for the Krško NPP. 11 ANFM2015 - Advances in Nuclear Fuel Management V Monday, March 30, 2015 Meeting Room: May ME2 Modeling Methods, Advances in Reactor Stability and Fuel Temperature Feedback for Steady-State and Transients Chair: Vincent Penkrot 4:20 PM 10 Modeling Methods for Tightly Packed Granular Fuel Abdalla Abou-Jaoude and Anna Erickson Nuclear and Radiological Engineering Program, Georgia Institute of Technology The paper investigates methods for modeling the neutronic behavior of fuels with Stochastic Granular Structures (SGS) to a high degree of fidelity and validates the models against equivalent homogenized cases. Granular fuels have recently been the subject of renewed attention due to their many attractive properties and their design flexibility. However, many fuels considered cannot reach high packing fractions, thus limiting their power density and heavy metal inventory. Developing and modeling fuels with higher packing fractions is therefore very desirable. An algorithm was developed to closely replicate tightly-packed structures within a cylindrical container. MCNP6 simulations were carried out using the obtained sphere coordinates and different metrics obtained were compared to homogenized models as well as models with simpler arrangements. The results were in good agreement and validate employing this SGS modeling method when a more exact representation of the microstructure is required. 4:45 PM 78 Simulation of CASL 3D HFP Fuel Assembly Benchmark Problem with On-the-Fly Doppler Broadening in MCNP6 Scott J. Wilderman and William R. Martin (1), Forrest B. Brown (2) 1) University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI, 2) Los Alamos National Laboratory, Los Alamos, NM An On-the-Fly Doppler broadening methodology has been applied in a neutronics simulation of a single fuel assembly (problem 6 of the CASL/VERA Core Physics Benchmark Problems) using MCNP6. HFP temperatures and densities were taken from results of a coupled neutronic-TH computation with the neutron transport code MPACT and the subchannel TH code COBRA-TF. An MCNP6 input file with over 13000 cells with independent temperatures and densities was constructed from a template input file for CASL/VERA problem 3 (3D HZP full assembly). OTF Doppler broadening coefficients for the 54 unique isotopes of the problem were generated using the routines provided in the MCNP6 distribution. HZP OTF MCNP6 results are compared with published benchmark results, and results for 3D HFP assembly simulations are compared with neutronics results from the coupled MPACT/COBRA-TF simulation. 5:10 PM 66 Development of COBRA-TF for Modeling of Full-Core, Reactor Operating Cycles Robert K. Salko and Travis Lange (1), Vefa Kucukboyaci, Yixing Sung (2), Scott Palmtag (3), Jess Gehin (1), Maria Avramova (4) 1) Oak Ridge National Laboratory, 2) Westinghouse Electric Company, 3) Core Physics, 4) The Pennsylvania State University CTF, the Pennsylvania State University version of COBRA-TF, has been adopted as the subchannel thermal hydraulic (T/H) capability in the core simulator being developed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). This has resulted in significant development efforts to expand the applicability of CTF to performing high-fidelity, full-core, multi-physics simulations. These efforts have focused on addressing CASL challenge problems for pressurized water reactors (PWRs), which include modeling of departure from nucleate boiling and CRUD induced power shift. Developments specific to full-core modeling capabilities include creation of a preprocessor utility for the user-friendly, rapid generation of pin-cell-resolved PWR models and implementation of a domain-decomposition parallelization of the code solution algorithm. In preparation for modeling CRUD growth phenomena, a coupling interface has been developed for CTF and the code has been incorporated into a multistate driver, which allows for modeling entire reactor operating cycles (i.e. years of operation). A simple CRUD modeling capability has been coupled to the code through this driver for capturing CRUD growth over these long operational periods. This paper presents an overview of these new features and shows results of a full-core, pin-cell resolved simulation of a Westinghouse 4-loop PWR core during a loss-of-flow transient as well as an initial coupled T/H-CRUD simulation of a 17 17 assembly during a 15-month reactor operation cycle. Meeting Room: Palmetto B ME3 Advanced or Extended Fuel Cycles and Economic analysis Chair: Craig Hove 4:20 PM 18 Updated Fuel Cycle Cost Model of the Fluoride-salt-cooled Hightemperature Reactor (FHR) Based on Neutronic Calculations Using MC Dancoff Factors Christopher Kingsbury and Bojan Petrovic Georgia Institute of Technology, Nuclear and Radiological Engineering, Atlanta, GA The Liquid Salt Cooled Reactor (LSCR), or Fluoride-salt High-temperature Reactor (FHR), is a type of Advanced High Temperature Reactor (AHTR), a generation IV reactor, currently under development by Oak Ridge National Laboratory (ORNL) for the U. S. Department of Energy, Office of Nuclear Energy’s Advanced Reactor Concept Program. The reactor design of 3400 MWt power employs graphite ‘planks’ filled with tristructuralisotropic (TRISO) fuel particles containing enriched uranium oxycarbide as fuel. Expected higher fabrication costs of this fuel type, combined with the low heavy metal loading that challenges cycle length, make an accurate evaluation of fuel cycle cost and characteristics very important. Our previous preliminary fuel cycle cost assessment employed multigroup (MG) burnup calculations in SCALE 6.1. However, the double heterogeneity of the fuel elements was not completely accounted for. The use of Monte Carlo based (MC) Dancoff factors allows correcting for these inaccuracies. Using the most recent fuel design specifications, appropriate MC Dancoff factors were calculated and applied. Use of these factors in MG depletion analysis yields corrected burnup data for use in a preliminary FCC model, which, in turn, informs the fuel design to minimize the cost of electricity. The results acquired and put forth by this research show the impact of the correction factors and identify an optimum fuel configuration under given assumptions. 12 ANFM2015 - Advances in Nuclear Fuel Management V Monday, March 30, 2015 4:45 PM 57 24-month PWR Fuel Cycles - Two Decades of AREVA Design and Operating Experience Craig Hove AREVA Inc., Lynchburg, Virginia, USA AREVA has over two decades of design and successful operational experience with 24-month nuclear fuel cycles for PWRs (and BWRs) in the USA. No safety, licensing or equipment problems have occurred. Twenty-four month cycles increase cycle capacity factors, but fuel cycle economics should not be neglected. The most economical 24month cycles with efficient use of uranium that minimize the metric kg U235 feed / GWd thermal energy production require assembly designs with heavy U-metal loadings, which is equivalent to low core power density in watts thermal per gram U-metal (w/g U-metal). For 24-month cycles, PWRs with high core power density (above 40 w/g Umetal) require very large inefficient feed batch sizes (i.e. large values of the metric kg U235 feed / GWd th) with degraded fuel cycle economics. The core power density can be lowered by switching to an assembly design with heavy U-metal loading. The AREVA (B&W) plants operating in the USA and the AREVA EPR plants being built around the world are all low power density cores and are thus very suited to economical 24-month cycles. For 24-month cycles, the UO2-Gd2O3 integral burnable neutron absorber is needed to control reactivity and peaking to avoid supplementary removable burnable poison components and to minimize gas pressure buildup in fuel rods. 5:10 PM 55 Economic Assessment of Accident Tolerant Fuel Cladding Options Nathan Andrews, Koroush Shirvan, Ed Pilat, Mujid S. Kazimi Massachusetts Institute of Technology, Cambridge MA If an accident tolerant fuel cladding is to replace the zirconium alloys, it will have to be economically viable. Four proposed materials are examined as cladding options: Stainless steel (SS), FeCrAl alloy, molybdenum (Mo) tri-layer composite and silicon carbide ceramic matrix composite (SiC CMC), each having its own development time and costs. The thickness of each cladding was assumed to reflect the strength of the material and its manufacturing limits. The UO2 enrichment savings or penalty was calculated for each cladding option relative to Zircaloy, given unit costs from recent market conditions. Based on this analysis, it was found that all options may end up requiring higher enrichment for the same fuel cycle length. SiC will likely be the least cost option. If the present value of avoiding a large reactor accident with a large radioactivity release is estimated using past experience for LWR large accidents, there is a definite net economic benefit relative to typical Zircaloy cladding only in using SiC CMC as a cladding material, when the ATF cladding is assumed capable of preventing reactor loss and radioactivity release in a Fukushima-type event. Because of the high enrichment costs relative to accident costs, there is only a marginal economic benefit in using SiC to prevent a core-only loss without radioactivity release (TMI-type) accident and a large economic loss using metallic ATF concepts. 5:35 PM 80 Nuclear Fuel Management Capacity Building Initiative for the Perspective of Introducing Nuclear Power in Morocco Bouhelal Oum Keltoum Higher National School of Mines of Rabat ENSMR Organization, Dept Industry Process, Rabat, Morocco This paper gives a broad picture of the Moroccan nuclear program development and the benefit of launching nuclear power in the framework of the national energy planning strategy and under the aegis of the IAEA. The paper describes the situation of the Moroccan energy sector and the current nuclear activities: although relevant efforts are continuously made to consolidate the infrastructures that aim to enhance nuclear knowledge and international cooperation, the challenge posed by the nuclear fuel cycle management for a newcomer country like Morocco remains today an important issue; better performing in building capacity is necessary and an example of upgrading disciplines related to the nuclear fuel cycle management of the nuclear education program taught in the existing universities and engineering schools is proposed. 13 ANFM2015 - Advances in Nuclear Fuel Management V Tuesday, March 31, 2015 Room: Sabal TA1 Core Analysis Tools for Fuel Management: Modeling and Validation - Part 1 Chair: Keith Drudy 8:00 AM 58 AP1000® PWR Startup Core Modeling and Simulation with VERA-CS F. Franceschini (1), A. T. Godfrey, S. Stimpson, T. Evans, B. Collins, J. C. Gehin, J. Turner (2), A. Graham, T. Downar (3) 1) Westinghouse Electric Co. LLC, Cranberry Township, PA, USA, 2) Oak Ridge National Laboratory, Oak Ridge, TN, 3) University of Michigan, Ann Arbore This paper describes the application of the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000®1 PWR. The AP1000 PWR features an advanced first core with radial and axial heterogeneities and at-power control rods insertion to perform the MSHIM™ advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. This paper focuses on the qualification efforts at hot zero power conditions, where Monte-Carlo reference solutions have been established. The comparison of both global core parameters (e.g. critical boron concentration, rod worth and reactivity coefficients) and finemesh fission rate spatial distribution indicate excellent numerical agreement between VERA-CS and the Monte-Carlo predictions across the simulations performed. 8:25 AM 63 Two-Dimensional BWR Core Analysis using Multi-Assembly CASMO5 and SIMULATE5 Rodolfo M. Ferrer, Joshua M. Hykes, Joel D. Rhodes III (1), Tamer Bahadir (2) 1) Studsvik Scandpower, Inc., Idaho Falls, ID 83404-3345, USA, 2) Studsvik Scandpower, Inc., Waltham, MA An analysis of a two-dimensional (2D) Boiling Water Reactor (BWR) using Multi-Assembly CASMO5 (C5) and SIMULATE5 (S5) is presented in this work. To model representative conditions of a BWR core depletion in 2D, certain approximations are introduced and a methodology presented that allow users to verify the accuracy of the nodal homogenized S5 model relative to a high-order, full-core transport results provided by multi-assembly C5. Previous assembly burnup and void history, as well as instantaneous conditions such as fuel temperature and void distribution, may be consistently modeled in both codes using the approach presented in this paper. Numerical results, which demonstrate the high level of accuracy achieved by the S5 nodal model relative to the reference fine-mesh transport solution obtained from C5, are presented for a BWR cycle depletion. 8:50 AM 1 Improvements in TIP and Gamma Scan Predictions in the next Generation GNF BWR Core Simulator AETNA02 James E. Banfield, Tatsuya Iwamoto, Jason Mann GE Hitachi Nuclear Energy (GEH)/Global Nuclear Fuel (GNF) Global Nuclear Fuel (GNF)’s next evolution of advanced Boiling Water Reactor (BWR) simulator is the LANCER02/AETNA02 lattice physics/BWR core simulator. A state-ofthe-art lattice physics model using two-dimensional Method Of Characteristics (MOC) from LANCER02 is coupled with a three-group semi-analytic nodal method for core flux solution in the AETNA02 core simulator, which is embedded within a flexible online core monitor system. AETNA02 includes a new model for in-core instrument response calculations accounting for intra-nodal power tilt based on MCNP experience which is able to improve the fidelity of predicted Traversing In-core Probe (TIP) signals, particularly for gamma TIP plants, as well as thermal TIP plants. This TIP response model is described and studied within the advanced lattice-physics and core-simulator system LANCER02/AETNA02. In this paper, several plants from the GNF database will be examined and compared to PANAC11 TIP response predictions. Substantial improvement over the previous TIP response model predictive capability is demonstrated in this paper. In addition, pin-by-pin gamma scan comparisons will be presented which also demonstrate the integral effect of both lattice physics and core physics modeling on fidelity compared to physical measurement. The impact of each new model in AETNA02 including the three-group flux solution, the new thermal hydraulic model, the multi-blade and blade depletion feedback, as well as the new TIP model will be studied separately as well as in combination to see the breakdown of improvement as well as the integral improvement. Accuracy in instrument response predictions is a critical part of the amount of adaption necessary and on the reduction of on-line/off-line biases and operational stability. Accuracy in the pin power predictions is closely related to uncertainties used in the establishment of thermal limits. Room: May TA2 Nodal and Lattice Physics Methods - Part 1 Chair: Scott Palmtag 8:00 AM 45 Improved PWR Radial Reflector Modeling With SIMULATE5 Tamer Bahadir Studsvik Scandpower, Inc., Waltham, MA The recent improvements implemented in Studsvik’s next generation code package CMS5, with CASMO5 and SIMULATE5, in modeling the PWR radial baffle/reflector is presented in this work. The shortcomings in the conventional approach of generating radial homogenized cross-sections and discontinuity factors from a 1D fuel/reflector transport calculation have been eliminated by re-computing the reflector node cross-sections and discontinuity factors in real core geometry by using the submesh calculation model in SIMULATE5. The submesh constants for the radial reflectors are generated from either a 1D fuel/reflector transport calculation or a multi-assembly core transport calculation. The effects of radial reflector modeling on core eigenvalue and assembly power predictions are demonstrated for the BEAVRS benchmark problem. 14 ANFM2015 - Advances in Nuclear Fuel Management V Tuesday, March 31, 2015 8:25 AM 20 Westinghouse Development and Customer Support for the New Core Analysis and Design System Vincent S. Penkrot, William A. Boyd, Baocheng Zhang, and Kevin T. Lasswell Westinghouse Electric Company, Cranberry Township, Pa Westinghouse Electric Company has developed its next-generation core analysis and monitoring software products. This new code set more tightly integrates the ANC core simulator and the BEACON™ Core Monitoring System*. Additionally, Westinghouse has developed a new system for nuclear data generation named NEXUS. Westinghouse recognized that the development of this new system, which replaces the ALPHAPHOENIX- ANC (APA) and BEACON version 6 systems, required not just a software development effort, but a complete utility transition effort including training and licensing support. This system is now in use by a number of US and international utilities. This paper will describe the features of the NEXUS system, updates to ANC in version 9, integration of ANC version 9 and BEACON version 7 as well as support tools used by Westinghouse to assist utility transition. 8:50 AM 21 Southern Nuclear’s Implementation of Westinghouse’s Next Generation Core Design Simulator and Core Monitoring Software Robin D. Jones (1), Gary T. Wolfram (2) 1) Southern Nuclear Operating Company, Birmingham, AL, 2) Westinghouse Electric Company, Rock Hill, SC The Southern Nuclear Operating Company (SNC) has been working with Westinghouse Electric Company (WEC or Westinghouse) on early implementation of a tightly integrated ANC (ANC9) core simulator and the BEACON™ Core Monitoring Software Version 7 (BEACON7) core monitoring system as well as the development of a new system for nuclear data generation named NEXUS. This paper will describe SNC efforts to support Westinghouse in the development, debugging, benchmarking and final implementation of the new methodologies at the Farley and Vogtle units. SNC has also developed ANC9 models for the Vogtle 3 and 4 cores. These models have the unique features of the integrated ANC9 system needed to support the modeling of the new Westinghouse AP1000 reactor design. In addition, this paper will describe some of the difficulties experienced in the long implementation process and outstanding issues. 9:15 AM 22 Finite Difference Method with Corrective Coupling Coefficient for Neutron Diffusion Calculation of Nuclear Reactor Core Analysis Jae-Seung Song and Jin Young Cho Korea Atomic Energy Research Institute, Yuseong-gu, Daejeon, Korea A finite difference method with a single radial mesh per fuel assembly, of which interface coupling coefficients are corrected using the information obtained from threedimensional whole core transport calculation, is proposed to solve the neutron diffusion equation for the nuclear reactor core. The nodal group constant for each mesh and the corrective interface coupling coefficient for each interface are tabulated as functions of the parameters affecting the neutron spectrum such as the soluble boron concentration, the fuel temperature and the moderator density. The method is tested through a core calculation for the pressurized water reactor, based on comparisons of the power distribution and the effective multiplication factor with those of a three-dimensional whole core transport calculation. Room: Palmetto B TA3 Advanced Fuel Management, Multi-dimensional Burnup Analysis, and Depletion Chair: Bojan Petrovic 8:00 AM 64 Methodology of the On-Line Fuel Management of Pebble Bed High Temperature Reactors Including Follow and Prediction Methods Bing Xia and Fu Li, Chunlin Wei, Jian Zhang, Jiong Guo Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Tsinghua University, Beijing, China The on-line fuel management is an essential feature of the pebble-bed high-temperature reactors (PB-HTRs), which is strongly coupled with the normal operation of the reactor. For the purpose of on-line analysis of the continuous shuffling scheme of numerous fuel pebbles, there are two kinds of calculations necessary for the PB-HTRs, i.e. the on-line follow calculations and prediction calculations. The keys of on-line follow calculations are the decoupling of pebble flow and steady-state neutronics calculations and the discretization of core layout and operation history. Based on the core status at a certain moment, a series of prediction calculations are implemented, and the best estimate of future operation scheme is made by polynomial interpolations. Both calculations are carried out by using the VSOP code system, and verified by the actual operation data of the HTR-10. 8:25 AM 25 Whole Core Analysis of Molten Salt Breeder Reactor Jinsu Park, Yongjin Jeong, and Deokjung Lee Ulsan National Institute of Science and Technology, Ulsan, Republic of Korea The simulation of whole core depletion and continuous reprocessing of Molten Salt Breeder Reactor (MSBR) has been performed. The MSBR model was built using MCNP6 and the depletion and reprocessing simulations were modelled using CINDER90 and PYTHON script. The PYTHON script was introduced to implement online reprocessing of molten-salt fuel and feeding of new fertile material with 3-day depletion intervals during the simulations. The simulation starts with the reference composition from the original ORNL MSBR design [3] and equilibrium compositions are searched through the depletion and reprocessing simulations for 1200 days. The MSBR whole core analysis was performed at the initial and equilibrium core conditions, for various reactor design parameters such as multiplication factors, neutron flux distributions, temperature coefficients, rod worths, and power distributions. The neutronic core characteristics was analyzed using four factor formula applied to the two zones of the core separately. 15 ANFM2015 - Advances in Nuclear Fuel Management V Tuesday, March 31, 2015 8:50 AM 49 Design of a Fast Breed/Burn Reactor Core Using the Deterministic Code KANEXT Roberto Lopez-Solis and Juan Luis Francois-Lacouture Universidad Nacional Autónoma de México, Facultad de Ingeniería, Morelos, México Fast breeder reactors are an interesting type of nuclear systems, due to, given the correct design conditions, they can generate more fissile fuel than they consume; nevertheless, in order for that breed fuel to be usable, it must be extracted from spent fuel an reprocessed before being used in the fabrication of new fuel. On the other hand, in a Breed/Burn reactor (B&B), bred plutonium is burned “in situ”, inside the core, just after being bred; this reduce costs and fuel proliferation by simplifying the fuel cycle. In this work, we present a B&B reactor design consisting of 210 active fuel assemblies plus 7 spaces for control rod assemblies. This core differs from most of the B&B reactors in its design that include a blanket zone in the center of the core; this is to take advantage of the population of fast and epithermal neutrons in the center of geometry, due to the fissions in adjacent zones. A satisfactory fuel reshuffling scheme was found in which the reactor operated for about 38 years, about 10 years more of what would have operated without any reshuffling scheme. Regarding the tool for the full core calculations, KANEXT code was used. Since this is a deterministic code, hardware needs were easy to satisfy and computational time was suitable for this kind of trial and error repetitive fuel reshuffling tests. 16 ANFM2015 - Advances in Nuclear Fuel Management V Tuesday, March 31, 2015 Room: Sabal TM1 Core Analysis Tools for Fuel Management: Modeling and Validation - Part 2 Chair: Bob StClair 10:00 15 Optimizing the In-Core Fuel Management of BWRs using Rosa L Gilli and P H Wakker NRG Utrechtseweg 310, Arnhem, The Netherlands The in-core fuel management of a nuclear reactor is a challenging task due to the virtually infinite number of possible loading patterns one could theoretically adopt. The ROSA (Reloading Optimization by Simulated Annealing) code is an optimization tool that has been successfully used in the last couple of decades to facilitate the core design of several Pressurized Water Reactors. In this paper we discuss the ongoing development of a version of ROSA capable of performing the core design of Boiling Water Reactors. We focus the discussion on the modifications of the neutronic module of the code that are needed to improve the accuracy when performing depletion calculations of BWRs. 10:25 69 Pellet-Cladding Mechanical Interaction Analyses Using VERA B. T. Mervin, M. L. Pytel, D. F. Hussey, and S. M. Hess Electric Power Research Institute, Palo Alto, CA A CASL Test Stand was launched in 2013 to evaluate VERA’s fuel performance component, BISON-CASL, as a state-of-the-art fuel performance code for PCI analysis by guiding it through a series of fuel performance progression problems. The progression problems are performed using 2D R-Z axisymmetric models and focus on examining the thermal and mechanical responses of the fuel and cladding to an imposed axially-varying power history. The progression begins with a constant axial power profile imposed during a single cycle ramp up to power followed by steady-state operation for a short length test rod and concludes with the most complex case studied by the Test Stand: a full-length fuel rod with an axially-varying power history containing a first cycle ramp to full power steady-state operation followed by a shutdown and a second-cycle ramp to full power. The evaluation of these progression problems is performed by comparing BISON-CASL results against results from the Falcon fuel rod performance code. The results of this comparison show that while differences exists in the thermomechanical responses between the two codes, the peak inside cladding surface hoop stress calculated by the two codes are within 0.5% of one another. 10:50 85 Physics-guided Coverage Mapping (PCM): A New Methodology for Model Validation Hany S. Abdel-Khalik (1) and Ayman I. Hawari (2) 1) School of Nuclear Engineering, Purdue University, West Lafayette, IN, 2) Department of Nuclear Engineering, North Carolina State University, Raleigh, NC This manuscript deals with a fundamental question in model validation: given a body of available experiments and an envisaged domain of reactor operating conditions (referred to as reactor application), how can one develop a quantitative approach that measures the portion of the prior uncertainties of the reactor application that is covered by the available experiments? Coverage here means that the uncertainties of the reactor application are originating from and behaving in exactly the same way as those observed at the experimental conditions. This approach is valuable as it provides a scientifically defendable criterion by which experimentally measured biases can be credibly extrapolated (i.e., mapped) to biases for the reactor applications. This manuscript introduces a novel approach, referred to as physics-guided coverage mapping (PCM) which provides a natural solution to this problem by relying on high fidelity physics simulation. We discuss the potential advantages of PCM over the methods of similarity indices, data assimilation, and model calibration commonly employed in the nuclear community. 11:15 23 A Method to Optimize Robust Core Design Performance Based on Design for Six Sigma (DFSS) Methodology Serkan Yilmaz Global Nuclear Fuel – Americas, Wilmington, NC This paper provides the detailed information about the methodology of newly developed Robustness Evaluation Module of ePrometheus. It summarizes the background and motivation of the work, Robust Core Design Description based on Design For Six Sigma (DFSS) methodology, design basis assumptions, historical simulation uncertainties in the reactor operations, development of robustness methodology and robustness metric for optimization response surfaces, opportunities and defect definition based on robustness performance and calculate statistical long-term process capability predictions based on DFSS methodology. A new ePrometheusTM prototype has been developed and incorporates this method as a modular functionality. 11:40 24 Optimizing the Outage Refueling Time with Shuffle Conscious Core Design Evaluation via ePROMETHEUS™ Serkan Yilmaz (1), John A. Elam (contributor) (2) 1) Global Nuclear Fuel – Americas, Wilmington, NC, 2) Nuclear Engineering Consultant, Leland, NC This paper provides the detailed information about the methodology of newly developed Robustness Evaluation Module of ePrometheus™. It summarizes the background and motivation of the work, Robust Core Design Description based on Design For Six Sigma (DFSS) methodology, design basis assumptions, historical simulation uncertainties in the reactor operations, development of robustness methodology and robustness metric for optimization response surfaces, opportunities and defect definition based on robustness performance and calculate statistical long-term process capability predictions based on DFSS methodology. A new ePrometheusTM prototype has been developed and incorporates this method as a modular functionality. 17 ANFM2015 - Advances in Nuclear Fuel Management V Tuesday, March 31, 2015 Room: May TM2 Nodal and Lattice Physics Methods - Part 2 Chair: Steve Bowman 10:00 50 Comparative Neutronics Analysis of DIMPLE S06 Benchmark Wonkyeong Kim, Jinsu Park, Deokjung Lee (1), Tomasz Kozlowski (2) 1) Ulsan National Institute of Science and Technology- UNIST, Ulsan, Republic of Korea, 2) University of Illinois, Urbana-Champaign, USA The DIMPLE S06 critical benchmark has been calculated using a direct modeling and a two-step modeling approaches. A direct modeling is performed with MCNP6, SERPENT2, CASMO4E, and TRITON/NEWT. A two-step modeling is performed by using the nodal code PARCS3.0 employing the homogenized two-group cross section which are generated through SERPENT2, CASMO4E and TRITON/NEWT. Detailed calculation models are developed in this paper and the parameters for a two-step modeling generated by each code are compared to each other. Finally, the eigenvalue for DIMPLE S06 are compared to each a direct and a two-step modeling. In this paper, the calculation result of two-step modeling agrees well with that of direct modeling. The result shows that the use of the assembly discontinuity factor in two-step calculation is essential to improve the relative error to the result of a direct modeling. A necessity of the assembly discontinuity factor is clearly proved by calculating the flux distribution for the homogenized assembly. 10:25 47 CASMO5 Analysis of NCA Tungsten Critical Experiments Joshua Hykes and Rodolfo Ferrer Studsvik Scandpower, Inc., Idaho Falls, ID This analysis of a series of critical experiments demonstrates the accuracy of CASMO5’s predicted fission rate distributions in the presence of tungsten gray control rods, which are inserted during operation of the AP1000 PWR. The RMS error of the computed fission rate range between 0.007 and 0.014 over the five core configurations, with a maximum absolute di erence of 0.025. This compares favorably with the quoted 2% standard deviation in the measurements. Comparison of identical 2D CASMO5 and MCNP6 models of the cores reveals excellent agreement for the core multiplication factors, within 100 pcm when using CASMO5’s 586 energy group structure. 10:50 2 Automated Reactor Records Evaluation Framework Jonatan Hejzlar and Frantisek Havluj Reactor Physics Department, UJV Rez, s.r.o. Plant operational data evaluation has a key role in the core physics code validation process. However, the amount of the data coming from the reactor operation is often so vast that it can be discouraging for the code developers to use it properly, often resulting in a reduction of the data set used which may easily introduce bias into the uncertainty quantification rendering the uncertainty results questionable. We present an elaborate, fully automated framework, which we have designed and implemented in our institute, for reactor records processing and its use for core physics code validation. Through a high level of automation this framework resolves the many difficulties in dealing with operational data giving easy and painless access to all the reactor records. This framework has been used for the validation of our ANDREA v2.0 core physics code before the Czech nuclear regulatory body. It has shown to be a powerful tool for comparing various code and library versions between themselves. It has also shown to be a tool useful for the validation of the received data itself. 11:15 82 CRANE: A New Scale Super-Sequence for Neutron Transport Calculations Congjian Wang and Hany S. Abdel-Khalik (1), Ugur Mertyurek (2) 1) School of Nuclear Engineering, Purdue University, West Lafayette, IN, 2) Oak Ridge National Laboratory, Oak Ridge, TN A new “super-sequence” called CRANE has been developed to automate the application of reduced order modeling (ROM) to reactor analysis calculations under the SCALE code environment. This new super-sequence is designed to support computationally intensive analyses that require repeated execution of flux solvers with variations in design parameters and nuclear data. This manuscript provides a brief overview of CRANE and demonstrates its applications to representative reactor physics calculations. Specifically, two ROM applications are demonstrated, the intersection subspace-based approach for uncertainty quantification which is intended to reduce the number of uncertainty sources in a conventional uncertainty analysis, and the exact-to-precision generalized perturbation theory methodology intended as a physics-based surrogate model to replace the flux solver, i.e., NEWT. Our overarching goal is to provide a prototypic ROM capability that allows users to further explore and investigate the benefits of using ROM methods in their respective domain and help guide further developments of the methodology and evolution of the tools. 18 ANFM2015 - Advances in Nuclear Fuel Management V Tuesday, March 31, 2015 Room: Palmetto B TM3 PANEL Discussion - "Adequacy of Methods for Nuclear Fuel Management” Chair: Jeffrey R. Secker 10:00 The panel discussion “Adequacy of Methods for Nuclear Fuel Management” will discuss the state of methods currently used for LWR reload fuel design and fuel management. Representative from all US vendors as well as a major software vendor are on the panel along with experts from academia and a national lab. Status of current methods and plans for future methods development will be discussed. Questions from the audience are welcomed. Panelists: Paul Turinsky, NC State University Jess Gehin, Oak Ridge National Lab Keith Drudy, Westinghouse Electric Co Tamer Bahadir, Studsvik Russ Stachowski, Global Nuclear Fuel Kevin Segard, Areva 19 ANFM2015 - Advances in Nuclear Fuel Management V Tuesday, March 31, 2015 Room: Sabal TE1 Thorium Cycles, MOX utilization, and Plutonium Disposition Chair: Ron Ellis 4:20 PM 19 Performance of Thoria Fuels and SiC Cladding for Burning of Plutonium in Pressurized Water Reactors Yanin Sukjai and Mujid S. Kazimi Massachusetts Institute of Technology, Cambridge, MA For burning of excess weapons plutonium in nuclear reactors, thorium is a better host fuel, since it does not generate new plutonium as uranium does, thus allowing faster depletion of the Pu stockpile. In this study, we compare the performance of two types of fuel materials (UO2-PuO2 & ThO2-PuO2) and two types of cladding (Zircaloy-4 & SiC) during irradiation, given a certain power history and axial peaking factors generated from full-core neutronic simulation. ThO2 has higher thermal conductivity and lower fission gas release rate than UO2. Thus, it is expected to enable better fuel performance during irradiation. Compared to zirconium based cladding, silicon carbide (SiC) has several desirable characteristics such as higher melting point, lower neutron absorption cross-section and better corrosion resistance. However, the low thermal conductivity of irradiated SiC and lack of creep down towards the fuel lead to a higher fuel temperature during irradiation. To reduce the high fuel temperature, the possibility of replacing the fuel-cladding gap bond material with lead bismuth eutectic (LBE) or high thermal conductivity porous foam is investigated. FRAPCON 3.4 was modified to allow the simulation of all these variants, and used to assess the best approach to Pu burning in PWRs. The results indicate that thoria reduces fission gas release compared to urania, thus reducing the internal fuel rod pressure. In addition, a foam bond material, such as alumina or graphite, will further reduce the temperature of the fuel with SiC cladding, and enable better fuel performance. 4:45 PM 9 The TRU-Incinerating Thorium RBWR Core Preliminary Design Phillip Gorman, Sandra Bogetic, Guanheng Zhang, Massimiliano Fratoni, Jasmina Vujic, and Ehud Greenspan University of Berkeley, California This study searches for the optimal design for the RBWR-TR – a reduced moderation BWR with a high transuranic (TRU) consumption rate. It is charged with LWR TRU and thorium and recycles all actinides unlimited number of times while discharging only fission products and trace amounts of actinides. This design is a variant of the Hitachi RBWR-TB2, which arranges its fuel in a hexagonal lattice, axially segregates seed and blanket regions, and fits within an ABWR pressure vessel. The RBWR-TR eliminates the internal axial blanket, eliminates absorbers from the upper reflector, and uses thorium rather than depleted uranium as the fertile makeup fuel. The design variables of the parametric studies include the fuel pitch-to-diameter ratio, number of fuel rods per assembly, length of each fuel section, TRU feed fraction, coolant flow rate, and fuel residence time. The sensitivity of the void feedback, cycle length, burnup, TRU consumption efficiency, shutdown margin, and critical power ratio to variation in each of the design variables were quantified to guide the design. A design is presented which incinerates TRU at a slightly higher rate per GWeY and discharged significantly less plutonium of a smaller fissile fraction than the reference ABR and RBWR-TB2 while meeting all the design constraints. However, due to significantly lower discharge burnup the RBWR cores require significantly larger reprocessing and fuel fabrication capacity per GWeY than the reference ABR. 5:10 PM 52 Comparison of Thorium and Uranium Fuel Performance in VVER-1000 Reactor Jan Frybort Czech Technical University in Prague, Department of Nuclear Reacors, Prague, Czech Republic Thorium can be considered as an alternative to uranium fuel. This analysis deals with ThO2 fuel for Pressurized-Water Reactor VVER-1000 and its comparison to UO2. Operational and spent fuel characteristics are considered. The main advantage of thorium is limited production of minor actinides and its greater abundance in nature. Problem of thorium utilization is absence of a fissile nuclide in the natural thorium, thus thorium fuel needs to be supported by an added fissile material. Addition of 233U and low-enriched uranium with 19.75 % enrichment is considered. The analysis is carried out using ANDREA nodal code with diffusion data prepared by HELIOS calculations. Room: Palmetto B TE3 PANEL Discussion - "Ongoing Fuel Performance and Fuel Cycle Issues - Driving to Zero" Chairs: Rob Schneider & Paul Cantonwine 4:20 PM The panel discussion “Ongoing Fuel Performance and Fuel Cycle Issues: Driving to Zero:” will cover a variety of operational issues for fuel in both BWR and PWR reactor designs such as fuel reliability, BWR channel distortion and the benefits of annual cycles to fuel cycle costs. The perspective of both BWR and PWR fuel vendors and experienced plant operators will be provided by the panelists. Panelists Rob Schneider, Global Nuclear Fuel Paul Cantonwine, Global Nuclear Fuel Brian Beebe, Westinghouse Jim Tusar, Exelon Bill Herwig, SCE&G 20 ANFM2015 - Advances in Nuclear Fuel Management V Wednesday, April 1, 2015 Room: Sabal WA1 Automated and Interactive Fuel Management Design and Optimization Tools - Part 1 Chair: Gerardo Grandi 8:00 AM 61 Recent Developments of the ROSA PWR Code and a Special Loading Pattern Design Application F.C.M. Verhagen, H.P.M. Gibcus, P.H. Wakker (1), D. Janin, M. Seidl (2) 1) NRG, Arnhem, The Netherlands, 2) E.ON Kernkraft GmbH, Hannover, Germany The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG’s loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators for almost two decades for improving their fuel management economics in a more and more constrained environment. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and other functionality. This paper outlines recent developments of the ROSA code system with a focus on the new full core version, DNBR-capability, and a special End Of Life (EOL) loading pattern design application. 8:25 AM 46 Detailed VIPRE Core Models from SIMULATE-3K Gerardo Grandi and Jerry Judd Studsvik Scandpower, Inc., Idaho Falls, ID In depth analysis of Reactivity Initiated Accidents (RIA) and plant transients in Pressurized Water Reactors (PWR), requires the integration of many different analytical tools. One such tool, the 3D nodal transient code SIMULATE-3K1 (S3K) has been coupled with the system codes RELAP5-3D,2 RELAP5-Mod3.33 and TRACE,4 to provide a bestestimate coupled code system for performing plant transient calculations with reactivity feedback from a detailed core model.5,6 More recently, S3K has been coupled with the fuel performance code ENIGMA.7 The combination of these two codes provides a powerful analytical tool for the analysis of RIA.8 In line with these previous developments, one would like to be able to have an in-depth analysis of the fuel assemblies Thermal-Hydraulic (TH) performance during RIA and plant transients. The first step in this direction is the interface between S3K and VIPRE.9 The purpose of this paper is to describe the status of the S3K/VIPRE interface and to show its application to a Rod Ejection Accident (REA) scenario. 8:50 AM 54 Designing Optimized Shuffles with SOSA P.H. Wakker, H.P.M. Gibcus and F.C.M. Verhagen NRG, Arnhem, The Netherlands SOSA is NRG’s software package for design and optimization of fuel shuffles. SOSA is focused on shortening the reload time of the reactor core. By reducing the movements of the refueling machine to a minimum, the reload time can be shortened by as much as 6-8 hours. This paper describes a few of the code’s features, such as the way to divide a shuffle into segments or phases, the approach to guarantee sufficient SDM, possibilities for spent fuel pool optimization and SOSA’s capability for generating move sheets. Finally, a couple of recently obtained results for different nuclear power plants are summarized. 9:15 AM 68 A New MIP Based Loading Pattern Search Tool Frank Popa Westinghouse Electric Company LLC, Cranberry Township, PA The Pearls™ loading pattern search tool has generated reactor core loading patterns with significant fuel cycle cost benefits over the years. This mixed integer linear programming (MILP) based tool has worked well for so called standard loading pattern searches. A new tool (TNT) is under development at Westinghouse that will build on the success of Pearls and on the once through cross section capability of NEXUS but will remove the limitations of Pearls. This will again be a MILP based method. TNT will handle the full panoply of objective functions and constraints. All PWR currently used burnable absorbers will be included. One particularly difficult aspect of loading pattern search is the choosing of the feed pattern. Two very different feed patterns may yield near optimum loading patterns once all the other decisions are made. An unusual feature of this new tool is that it systematically and exhaustively analyzes all feed patterns within the MILP framework. The expectation is then that the final loading patterns will be global optima within the accuracy of the associated licensed reactor core neutron flux solver. Room: May WA2 Advanced Fuel Assembly and Burnable Absorber Designs Chair: Christian Malm 8:00 AM 11 Neutronic and Economic Evaluation of Accident Tolerant Fuel Concepts for Light Water Reactors Ian Younker (1), Massimiliano Fratoni (2) 1) The Pennsylvania State University, University Park, PA, 2) University of California, Berkeley, CA Ceramic clad coatings and alternative cladding materials are a few of many accident tolerant fuel (ATF) concepts. Each concept looks to reduce the amount of zirconiumalloy cladding available for reaction with high temperature steam. In order to be implemented into current and future light water reactors (LWR), ATF concepts must provide enhanced neutronic and economic performance over conventional Zircaloy-UO2 fuel. This study used a single assembly pressurized water reactor (PWR) model to investigate reactivity drop, cycle length penalty, enrichment compensation, and reactivity coefficients, and a fuel cost model to understand economic performance. Findings show a desirable thickness of 10-30 μm for ceramic clad coatings to reduced neutronic economic penalties. For alternative cladding materials, SiC accommodates thicker cladding while other alloys, due to large neutronic penalties, require thin tubes and/or higher enrichment. 21 ANFM2015 - Advances in Nuclear Fuel Management V Wednesday, April 1, 2015 8:25 AM 59 A Full Core Integral Fuel Performance Assessment of SiC Cladding Alexander J. Mieloszyk, Ronald Gil, Koroush Shirvan, Mujid S. Kazimi Massachusetts Institute of Technology, Center for Advanced Nuclear Energy Systems, Cambridge, MA To better understand the implications of using SiC cladding in a commercial reactor, a framework has been developed to evaluate the fuel performance of a large fraction of the fuel rods in a typical PWR core. This framework makes use of the newly developed RedTail fuel performance code and the CASMO-SIMULATE reactor physics suite. Applying current component specific material properties, this framework is utilized to assess a three-layer SiC cladding design and compare it to the performance of zirconium-based clad fuel under the same conditions. This assessment reveals higher fuel temperatures and plenum pressures associated with the SiC cladding, similar to those observed in previous single pin analyses. Of note, however, is the observation that the SiC cladding stresses increase significantly during reactor shutdowns due to the presence of radial gradients of swelling (growth) strain. Additionally, the hottest and most burnt fuel rods do not present the most challenging conditions to the SiC clad fuel. This implies that a capability to analyze the fuel performance of an entire core is necessary to find the expected cladding failure risk associated with the deployment of various SiC cladding designs. 8:50 AM 53 Accident Tolerant Fuel and Resulting Fuel Efficiency Improvements Jeffrey Secker, Fausto Franceschini, and Sumit Ray Westinghouse Electric Company, LLC, Cranberry Township, PA Fuel designs using advanced, accident tolerant fuel materials can improve fuel efficiency and extend fuel management capability in addition to improving safety margins for LWRs. The use of SiC cladding material can reduce fuel cycle costs by about 2% if it can be manufactured to the current thickness of zirconium alloy based cladding in use in PWRs today. The increased pellet densities associated with the higher density U3Si2 or UN material also can reduce fuel costs by an additional 4-6% beyond the SiC cost reduction for 18 month cycles or 8-11% for 24 month cycles. Because of the increased density, the use of these materials also extends the energy output and cycle length capability for PWR fuel assemblies while remaining below the 5 w/o enrichment limit for commercial fuel and can make 24 month cycle operation economical for today’s uprated, high power density PWRs. Room: Palmetto B WA3 PANEL Discussion - "CASL: Consortium for the Advanced Simulation of Light Water Reactors" Chairs: Paul Turinsky (NC State University), Rose Montgomery (Tennessee Valley Authority) 8:00 AM The Consortium for Advanced Simulation of Light Water Reactors (CASL) is an Energy Innovation Hub established by the US Department of Energy in 2010 to advance the development and application of modeling and simulation technologies for nuclear reactors. CASL’s mission is to provide a step change in computational capabilities to the nuclear energy industry—one that enables more accurate prediction of the key phenomena defining the operational and safety performance of light water reactors (LWRs). Through CASL, experts from national laboratories, universities, and industry are developing and deploying CASL’s Virtual Environment for Reactor Applications (VERA), a “virtual reactor” designed to accurately simulate the physical processes inside a reactor at unprecedented levels of detail. These processes include neutron transport, thermal hydraulics, nuclear fuel performance, and corrosion and surface chemistry. VERA relies on the latest science-based physical models for nuclear reactor phenomena, advanced numerical methods for solution of these models, modern computational science and engineering techniques for imparting these methods into the VERA software, tools for estimating uncertainties and sensitivities of the VERA simulations, and validation against data from operating reactors and other pertinent experiments. More information is available at www.casl.gov. 8am – 8:15 Introduction to the panel discussion, Paul Turinsky 8:15-8:45 VERA Core Simulator, Scott Palmtag 9:00-9:30 VERA Neutronics, Scott Palmtag 10:00-10:35 VERA Thermal-Hydraulics & Chemisty, Bob Salko 10:45-11:15 Fuel Performance, Brenden Mervin 11:30-11:50 VERA Uncertainty Quantification and Validation Activities, Paul Turinsky Between each session the speakers will be available for Q&A 22 ANFM2015 - Advances in Nuclear Fuel Management V Wednesday, April 1, 2015 Room: Sabal WM1 Special Session - IAEA Presentation: Beyond 5% Enriched UO2 & ATF Progress. Chair: Victor Inozemtsev 10:00 Objective of the Special Session is to inform interested participants about the IAEA activities in the area of Fuel Engineering that are relevant to the scope of the ANFM. Particular attention with be paid and invitation extended to planned Technical Meeting “Beyond 5% enrichment limit for LWR: perspectives and problems” (12-16 October 2015, Vienna) and Coordinated Research Project “Analysis of options and experimental examination of fuels for water-cooled reactors with increased accident tolerance” (open for proposals, first meeting on 14-18 September 2015). Room: May WM2 Management, Design, and Operation Issues of Advanced Reactor Fuels, Practical Design Constraints, and Advances in On-Line Core Monitoring Chair: Tomaz Kozlowski 10:00 70 Multiphysics PWR Modeling Including Crud Induced Power Shift (CIPS) and Crud Induced Localized Corrosion (CILC) Andrew Petrarca, Jeffrey Secker and Michael Krammen Westinghouse Electric Company, Nuclear Fuel, Hopkins, SC The goal of the DOE’s Consortium for Advanced Simulation of Light Water Reactors (CASL) is to develop advanced multi-physics methods to improve reactor safety, reduce waste generation, and enable increased generation of carbon-free nuclear power. CASL is organized to solve problems that challenge operating PWR’s to meet the DOE goals, such as crud deposits on fuel, grid to rod fretting, fuel assembly distortion, and pelletclad interaction (PCI) that can lead to breaches in PWR fuel cladding. As an initial step in establishing multi-physics models for PWR crud deposition, the Westinghouse neutronics code ANC and thermal-hydraulics code VIPRE-W were linked with the EPRI crud and chemistry code BOA3.0 to predict fuel crud deposition. Westinghouse then upgraded the coupled package to make use of EPRI’s latest chemistry code, BOA3.1. A plant which experienced CIPS during an operating cycle was modeled for this analysis. 10:25 86 I2S-LWR Fuel Management Options for an 18-Month Cycle Length D. Salazar, F. Franceschini, P. Ferroni (1), B. Petrovic (2) 1) Westinghouse Electric Company LLC, Cranberry Township, PA, USA, 2) Nuclear and Radiological Engineering, Georgia Tech, Atlanta, GA, USA This paper presents the fuel management options developed for the Integral Inherently Safe LWR (I2S-LWR). The I2S-LWR is a reactor concept of a ~1,000 MWe (2,850 MWt) integral PWR with inherent safety features. The baseline core configuration contains 121 fuel assemblies with a 19×19 square lattice and 144-in active fuel height. The baseline fuel choice is U3Si2 in advanced FeCrAl-type steel cladding, which is envisioned to enhance accident tolerance but is detrimental to neutron economy. SiC cladding is also under consideration as it can foster further improvements in accident tolerance with excellent neutron economy. Standard UO2/Zr fuel is under investigation as an option for accelerated deployment. The performance of these three fuels, U3Si2/FeCrAl, U3Si2/SiC and UO2/Zr, is examined and compared in this paper; the focus is on fuel management and fuel cycle cost aspects for the I2S-LWR core at the equilibrium cycle with an 18-mo cycle length. 10:50 73 Overview of New MHI Online Core Monitoring System VISION Yuki Takemoto, Kazuki Kirimura, Naoko Iida, Shinya Kosaka, Hideki Matsumoto Nuclear Energy System Division, Mitsubishi Heavy Industries, Ltd., Hyogo-ku, Kobe, Hyogo, Japan Mitsubishi Heavy Industries, Ltd (MHI) has about 20-year experiences of supplying core monitoring systems and their maintenance for Japanese PWR utilities. MHI has developed a new online core monitoring system (OCMS) VISION. VISION is based on GARDEL-PWR provided by Studsvik Scandpower AB and has been improved to enhance the user-friendliness with MHI’s PWR core design experiences. Main features of VISION are automated core monitoring, reactivity management, core management and guidance of operation planning. Generally, online core simulator is built in a core monitoring system, which performs core calculation periodically to predict 3D power distribution with plant signals. Their information is displayed with the graphical user interface to support operational management. Usually core simulator installed in an OCMS is not replaced so often, because replacement of core simulator is not so easy for the customers. However, there would be a need to use different type of simulator, when core designer or fuel vendor is changed for example. Therefore, the multiple core simulator function has been developed to support easy core simulator replacement in VISION. As a core simulator engine of VISION, users can alternatively select between MHI’s new 3D core design code COSMO-S and Studsvik Scandpower’s core design code SIMULATE-3 for each cycles using the multiple core simulator function. Then, this paper describes the VISION’s overview including developed functions. 23 ANFM2015 - Advances in Nuclear Fuel Management V Wednesday, April 1, 2015 11:15 77 SMR Fuel Cycle Optimization and Control Rod Depletion Using Nestle and LWROPT Keith E. Ottinger, P. Eric Collins, Nicholas P. Luciano, and G. Ivan Maldonado The University of Tennessee, Department of Nuclear Engineering, Pasqua Engineering Building, Knoxville, TN The multi-cycle BWR fuel cycle optimization code BWROPT has been generalized to handle PWRs and SMRs and renamed LWROpt (Light Water Reactor Optimizer) and an eighth core symmetric shuffle option has also been implemented. The new features of the optimizer are tested using a test case based on an SMR model previously developed manually. Also, preliminary tests of a spatially-dependent and movable-region isotopic tracking feature under development in NESTLE are illustrated with the ultimate goal of assessing control rod depletion for very long SMR rodded cycles. 11:40 88 Deterministic Methods for PWR Fuel Loading Optimization Fariz Abdul Rahman and John C. Lee (1), Fausto Franceschini (2) 1) University of Michigan, Ann Arbor, 2) Westinghouse Electric Company LLC, Cranberry Township, PA, USA We have developed a multi-control fuel loading optimization code for pressurized water reactors (PWRs) based on deterministic methods. The objective is to flatten the fuel burnup profile, which maximizes overall energy production. The optimal control problem is formulated using the method of Lagrange multipliers and the direct adjoining approach for treatment of the inequality power peaking constraint. The optimality conditions are derived for a multi-dimensional multi-group optimal control problem via calculus of variations. Due to the Hamiltonian having a linear control, our optimal control problem is solved using the gradient method to minimize the Hamiltonian and a Newton step formulation to obtain the optimal control. We are able to satisfy the power peaking constraint during depletion with the control at beginning of cycle (BOC) by building the proper burnup path forward in time and utilizing the adjoint burnup to propagate the information back to the BOC. Our test result for the multi-control fuel loading design is able to achieve the same fuel cycle length as the AP600 first cycle loading with an average reduction of 0.02 wt% 235U enrichment. 24 ANFM 2015 - Advances in Nuclear Fuel Management V Wednesday, April 1, 2015 CASL VERA Workshop Room: Palmetto B Leaders: Dr. Mike Doster (1), Ms. Rose Montgomery (2) 1) North Carolina State University, 2) Tennessee Valley Authority Changes to Page 26, CASL VERA Workshop The Consortium for Advanced Simulation of LWRs (CASL) is developing a suite of tools for high fidelity coupled physics simulations of currently operating pressurized waterDoster reactors (PWRs). This Montgomery workshop will (2) provide an introduction to the CASL Virtual Environment for Leaders: Dr. Mike (1), Ms. Rose Reactor Applications (VERA). VERA incorporates science-based models, state-of-the-art numerical methods, modern computational sci1) North Carolina State University, 2) Tennessee Valley Authority ence and engineering practices, and uncertainty quantification tools to provide flexible simulation tools that span the range from atomistic to engineering scales. No change to the existing description text. More information on CASL and VERA are available at www.casl.gov. Currently, the agenda looks like this: Participants will explore the VERA Core Simulator and develop VERA input to run several simple cases. The class size is limited and participants must register in advance. Also, attendance in the CASL technical session is a prerequisite for the workshop. Participants in the workshop will be required to supply information for approval consistent with U.S. export control requirements, and should bring a laptop that has the necessary connect Pleasesoftware modify to it this way:to an external machine via ssh (e.g., Putty, No Machine) installed. Time slot 1 – 1:30pm 1:30pm 1:45pm 2:15pm 2:45 BREAK 3pm 3:45pm 4:15pm 4:45pm Topic Introductions, logistics Training Packets Machines & Login Introduction to VERA Common Input Quick Start Tutorial #1 – simple fuel rod cell (2D) Quick Start Tutorial #2 – 17x17 2D lattice (HFP, BOL, mid-‐plane) Speaker Mike Doster Troy Eckleberry Quick Start Tutorial #3 -‐ single 3D fuel assembly (includes T-‐H feedback and depletion) Quick Start Tutorial #4 -‐ 2D full core using quarter core symmetry (mid-‐plane, HFP, BOL) Demonstration – 3D full core with feedback and depletion. Closing, Feedback forms, RSA token return Bob Salko 25 Andrew Godfrey Andrew Godfrey Andrew Godfrey Andrew Godfrey Andrew Godfrey Mike Doster ANFM 2015 - Advances in Nuclear Fuel Management V March 29 - April 1, 2015 Notes 26 PROGRAM AT-A-GLANCE March 29 - April 1, 2015 SUNDAY 6:00 -9:00 PM Registration open 3:00 to 5:00 PM @ Palmetto Landing Welcome Reception - Shore House – Sponsored by Global Nuclear Fuel MONDAY Registration open 7:15 - 8:00 AM @ Palmetto Landing 7:00 - 8:00 AM Breakfast - Palmetto C Palmetto Ballrooms A & B 8:00 - 9:40 AM 9:40 - 10:00 AM 10:00 AM - 12:05 PM 12:05 - 4:00 PM 4:00 - 4:20 PM 4:20 - 6:00 PM 7:00 -10:00 PM Opening Plenary Session Coffee Break – Palmetto Landing – Sponsored by SC Electric & Gas Sabal May MM1 New Modeling Concepts, Reactivity Control, MM2 Innovative Core Loading, Reload Design, and Generation of Cross Section Libraries and Whole Core Licensing Transport Calculations Palmetto B MM3 PANEL Discussion - "Small Modular Reactors: Challenges and Proposed Solutions to Successful Deployment" Chair: Hany Abdel-Khalik Chair: Andrew Worrall Chair: Rodolfo Ferrer Lunch Break Coffee Break – Palmetto Landing – Sponsored by SC Electric & Gas ME1 Error Quantification of Core Simulation Capabilities, ME2 Modeling Methods, Advances in Reactor Stability and Core Follow Data to Enhance Core Simulation Fidelity, Fuel Temperature Feedback for Steady-State and Utilization of Zero Power Physics Tests Transients ME3 Advanced or Extended Fuel Cycles and Economic analysis Chair: Fausto Franceschini Chair: Craig Hove Chair: Vincent Penkrot Dinner – Shore House – Sponsored by Southern Nuclear Company • Attendance only by purchase of a ticket TUESDAY Registration open 7:15 - 8:00 AM @ Palmetto Landing 7:00 - 8:00 AM Breakfast - Palmetto C 8:00 - 9:40 AM Sabal TA1 Core Analysis Tools for Fuel Management: Modeling and Validation - Part 1 May TA2 Nodal and Lattice Physics Methods - Part 1 Palmetto B TA3 Advanced Fuel Management, Multi-dimensional Burnup Analysis, and Depletion Chair: Keith Drudy Chair: Scott Palmtag Chair: Bojan Petrovic 9:40 - 10:00 AM 10:00 AM - 12:05 PM 12:05 - 4:00 PM 4:00 - 4:20 PM 4:20 - 6:00 PM 7:00 -10:00 PM Coffee Break – Palmetto Landing – Sponsored by SC Electric & Gas TM1 Core Analysis Tools for Fuel Management: Modeling TM2 Nodal and Lattice Physics Methods - Part 2 and Validation - Part 2 TM3 PANEL Discussion - "Adequacy of Methods for Nuclear Fuel Management” Chair: Bob StClair Chair: Jeffrey R. Secker Chair: Steve Bowman Lunch Break Coffee Break – Palmetto Landing – Sponsored by SC Electric & Gas TE1 Thorium Cycles, MOX utilization, and Plutonium Disposition TE3 PANEL Discussion - "Ongoing Fuel Performance and Fuel Cycle Issues - Driving to Zero" Chair: Ron Ellis Chairs: Rob Schneider & Paul Cantonwine Banquet – Palmetto Ballrooms B & C – Sponsored by Westinghouse • Attendance only by purchase of a ticket WEDNESDAY Registration open 7:15 - 8:00 AM @ Palmetto Landing 7:00 - 8:00 AM Breakfast - Palmetto C 8:00 - 9:40 AM Sabal May WA1 Automated and Interactive Fuel Management Design WA2 Advanced Fuel Assembly and Burnable Absorber and Optimization Tools - Part 1 Designs Palmetto B WA3 PANEL Discussion - "CASL: Consortium for the Advanced Simulation of Light Water Reactors" Chair: Gerardo Grandi Chairs: Paul Turinsky and Rose Montgomery 9:40 - 10:00 AM 10:00 AM - 12:05 PM Chair: Christian Malm Coffee Break – Palmetto Landing – Sponsored by SC Electric & Gas WM1 Special Session - IAEA Presentation: Beyond 5% WM2 Management, Design, and Operation Issues of Enriched UO2 & ATF Progress. Advanced Reactor Fuels, Practical Design Constraints, and Advances in On-Line Core Monitoring Chair: Victor Inozemtsev WA3 continued… Chair: Tomaz Kozlowski 12:05 PM Main Conference concluded 12:00 - 1:00 PM Lunch for workshop attendees only - Edisto 1:00 - 2:45 PM WORKSHOP: CASL VERA Attendance restricted to those registered & approved 2:45 - 3:00 PM Coffee Break CASL VERA Workshop continued 3:00 - 5:00 PM 5:00:00 PM Meeting Adjourned
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