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