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.
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ANFM2015 - Advances in Nuclear Fuel Management V
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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.
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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.
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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.
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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.
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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.
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ANFM2015 - Advances in Nuclear Fuel Management V
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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.
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