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U.S. Status of Fast ReactorResearch and Technology
Robert HillArgonne National LaboratoryARC National Technical Director
June 21, 2012
2
Advanced Reactor Concepts (ARC)
The mission is to develop and refine future reactor concepts that could dramatically improve nuclear energy performan ce (e.g., sustainability, economics, safety, proliferation re sistance)
The strategic approach is to:� Tackle key R&D needs for promising concepts
– Fast reactors for fuel cycle missions– Fluoride salt cooled thermal reactor for high-temperature missions
� Develop innovative technology features with potenti al benefits to many concepts
� Utilize international collaborations to leverage an d expand R&D investments– Continuation of multi-lateral Generation-IV R&D Projects– Investment in strategic bilateral or trilateral partnerships
� Stimulate ideas for transformational reactor concep ts
3
Advanced Reactor ConceptsOrganizational Structure
ARC is organized into several technical areas:� Management and Integration ( Hill-ANL)� Fast Reactor Concepts (Grandy-ANL)� Thermal Reactor Concepts (Holcomb-ORNL)� Energy Conversion Technology (Rochau-SNL)� Nuclear Data (Hill-INL)� Generation-IV International Support (Connell-INL)� Transformational Concepts
Other DOE-NE R&D initiatives include advanced react or applications� Modeling and Simulation (NE-4)� Transmutation Fuels (NE-FCT)� Related University Program (NEUP) Contributions
Advanced Fast Reactor Concept Development
4
Concept development studies have several important purposes� Identify the key R&D needs and challenges � Pursue fundamental understanding of technical utili zation of
advanced technology options in integrated reactor s ystem– Confirm feasibility of innovative features– Estimate the impacts and benefits to prioritize new features
� Evaluate broad range of technology options– Different configurations and power sizes (e.g., SMR concepts)– Different coolant options, fuels, etc.
Thus, this work guides the R&D directions by screen ing the idea space with favorable technology options promoted to specific R&D technical areas
Concept Development Studies
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Fuel Cycle Mission:Fast Spectrum Reactors
� For closed fuel cycle options, must develop and dem onstrate recycle reactor technology– Significant waste management and resource extension benefits– Fast reactor needed for final transmutation system
� For future fast reactor technology, the key researc h focus is capital cost reduction (i.e., major commodity reduct ions or efficient electricity generation) through– Improved design approach (e.g., compact configuration)– Advanced technologies (e.g., materials, energy conversion)– Advanced simulation for optimized design
� A second research focus is assurance of safety to p romote design simplification and licensing
� A third, related focus is high system reliability
Fast Reactor Advanced Concept Studies (examples)
� Areas previously investigated include:
� Impact of Advanced Materials –potential for stronger materials reducing plant commodity usage
� Impact of Increasing Core Outlet Temperature – increased power output due to increased efficiency
� Advanced System and Components include:
– Compact fuel handling mechanisms– Advanced Balance of Plant Systems– Vented Fuel– Ultra-Long-lived fast reactor cores– Integrated primary purification
systems– Advanced heat exchanger
technology options
Fast Reactor Advanced Concept Studies (examples)
� FY11 and FY12 work focused on SMR (~100 MWe) concepts: unique features include long-lived core and fuel shuffling strategies
� Developed compact fuel handling mechanism:
– Single rotating plug configuration– Pantograph design– Offset from center
� Detailed analysis and design options for core restraint
– A key feature for inherent safety reactivity feedbacks
– NUBOW bowing analysis code recovered and refined
– Limited free bow design� AFR-100 safety analyses conducted
– Verify inherent safety in double fault transients
QR
NP
TLP
ACLP
B
C
Cor
ere
gion
TLP restraintring
ACLP
a. b. c.
AFR-100 Upper Internal Structure
Recent work to develop compact reactor configuration:
� Upper Internal Structure– Very complicated structure within
the primary plant– Made from 316SS and clad with
Alloy 718 for thermal fatigue resistance
– UIS is lifted by a lifting drive mechanism to clear some systems and components before and during rotation
– UIS is keyed into the core barrel structure to prevent lateral movement during seismic events
– Advanced Fuel Handling System is integrated with the UIS
March 20129
Mechanisms Engineering Facility
� The facility will test small and intermediate-scale sodium components, examples include:
– Components for advance fuel handling systems such as grippers, spline shafts, universal joints, bearings, etc.
– Instrumentation including detectors for rapid detection of impurities and improved methods for sodium level measurement
– In-service inspection and repair technologies
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Fast Reactor Advanced Concept Studies - Thorium fuelled AFR -100 core
� Purpose - The feasibility study of fuel changes from U-based fuel to Th-based fuel in a small SFR
� Various approaches with Th-TRU:1) Conventional 4-batches operation with Th-TRU (18.6wt%). No
U-Zr startup core.2) Progressive replacement of U-Zr assemblies with Th-TRU (14.4-
22wt%). No equilibrium operation3) Progressive replacement of U-Zr assemblies with Th-TRU
(18.6wt%). 4-batches equilibrium strategy
� Different performance with various pros and cons (p eak power; reactivity swing; discharge burnup…)
Characteristic Unit (1) (2) (3)
Average TRU wt% wt% 18.60% 18.65% 18.60%
Burnup reactivity swing %∆k/k 0.73 1.3 1.25
Average power density W/cm3 58.3
Peak power density W/cm3 110.3 112.7 114.8
Radial power peaking factor
- 1.32 1.36 1.45
Average/Peak BU – U-Zr
% n.a.7.4%/ 12%/
11.9% 22.2%
Average/Peak BU –Th-TRU
%10.2%/ 7.8%/ 10.2%/
16.2% 17.4% 16.2%
Burnup distribution at EOEC for (1)
1.000
1.005
1.010
1.015
1.020
0 5000 10000 15000 20000
k_
eff
EFPD
(1) Conventional; 4-batches equilibrium
(2) Progressive replacement; no equilibrium
(3) Progressive replacement; 4-batches equilibrium
Information Recovery Efforts
� Another aspect of the concept development scope involves information recovery efforts
– Recovery of Information at FFTF (PNNL)– Recovery of Information at EBR-II (INL)– TREAT information recovery (ANL)– Retrieval of fast reactor information from OSTI
� Some key activities that have been pursued include:– NUBOW code recovery – supports core restraint design– SWAAM (sodium-water interaction) code recovery
• Upgraded to include sodium-CO2 interactions– FFTF Transient Testing and Startup Physics Testing Data– FFTF Design Standards and Procurement Specifications for
major FFTF systems and components
Fast Reactor Safety
13
Fast Reactor Safety and LicensingR&D Tasks
� International Passive Safety Evaluations– IAEA-coordinated projects on SFR Passive Safety (Monju and Phenix benchmarks)– Benchmark Analysis EBR-II Shutdown Heat Removal Tests (starting in FY12)
• U.S. contribution as the next IAEA-coordinated research project (CRP)• Largest CRP in IAEA history with more than 20 organizations in 12 countries
� Database Development for Safety Model Validations– Archiving of data for EBR-II/FFTF/TREAT Tests and SFR fuels and materials– FY12 marks successful completion of EBR-II and TREAT Test Databases
� Safety Technology Assessment and Regulatory Develop ment Plan– Evaluation of the status of the existing technology base, identifying where gaps exists
and additional effort is required– Regulatory Development Plan is finalized to help identify future R&D focus
� International collaborations– Gen-IV International Forum (GIF) SFR Safety and Operations PMB– GIF SFR Safety Design Criteria Task Force– U.S.-China Safety Code Collaboration– U.S./Japan/France bilateral and trilateral collaborations (Safety STCs).
Monju Benchmark: Objectives
� IAEA coordinated research project for benchmark analysis of Monju plant turbine trip test
– Analysis of sodium circulation in the upper plenum of Monju for validation of the 3-D CFD models
– Coolant mixing and temperature distributions in the hot pool for evaluation of thermal stratification
– Assessment of sharp temporal and spatial temp. gradients in reactor vessel and upper core structures
– Other benchmark participants include JAEA/Univ. of Fukui, CEA, KAERI, IPPE, IGCAR, and CIAE
Monju: MOX-fueled, loop-type prototype sodium-cooled fast reactor with three primary coolant circuits producing 714 MW-thermal (280 MW-electric) energy
16
Importance of Bypass Flow through Inner Barrel Holes
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SFR Licensing Gap Analysis
• Design options• Event classes• Event sequence• Systems or components • Fuel, core, primary, vessel
Gap Analysis Coordination tasks• Establish objectives, high level criteria • Establish structure, sequence• Define system design options, safety relevant features • Define panel topic areas, panel objectives, membership
Fuels and Materials Panel(Chair: L. Walters)
Accident Sequence Panel(Chair: J. Sackett)
Computer Codes and Models Panel
(Chair: R. Schmidt)
Sodium Phenomena Panel(Chair: M. Corradini)
• Design options• Core support• Control rods, internals• Vessel, piping, •Intermediate heat exchanger
• Na events• Event conditions• Fires, aerosols, radionuclide chemistry
• Accident analysis codes•Neutronics - cross sections, reactivity coefficients, temperature dependence, codes •TH – convective flow, heat transfer, codes
Source Term Panel(Chair:. D. Powers)
•Radionuclide release to primary system –fuel types, FP chemistry Na, TH transport, vessel•Radionuclide release to environment secondary transport, chemistry, aerosol transport
� All five of the PIRT-like gap analysis panels compl eted� In FY12, results of the five groups are being integ rated and prioritized by an
experts panel with contributions from the chairs of five gap analyses� A “Regulatory Development Plan” for identified gaps is key FY12 deliverable
• Will help guide FY13+ R&D priorities in Fast Reactor Safety and Licensing Technical area
Summary of SFR Safety and Licensing Research Plan Recommendations
• Documentation of safety related codes and experiments risk being lost• Piecemeal and underfunded efforts will lead to lost information which may need to be reproduced in
the future
Coordinated Knowledge Management and Preservation E ffort
• Adequate stewardship and documentation of U.S. safety related codes required for licensing (e.g., LIFE-Metal)
• Modernization of U.S. Codes to satisfy current licensing needs• Code (e.g., SAS4A) improvements related to seismic response of the entire SFR system will be
required post-Fukushima • Probabilistic safety analysis of containment response capabilities need to be developed for SFRs
within the U.S. (i.e., incorporation of sodium phenomena into MELCOR)
Improvements to U.S. safety related codes
• Ensures that future testing capabilities are not lost in budget conscious environments• Identify testing to address phenomenological uncertainties which could be performed to maintain
facilities
Continued U.S. experimental facility utilization, ev en if on a small scale
• The current process makes removing AT designations on documents which no longer need to be protected extremely difficult.
• The U.S. NRC is not set up to handle AT documents.
Treatment of the Applied Technology (AT) designatio n must be streamlined
18
Fast Reactor Advanced Materials
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Fast Reactor Advanced Materials Consists of Several Key Areas
� Advanced Alloy Development: – Charged with procurement of alloys, defining and testing material
processing, and acquiring basic properties.� Advanced Alloy Testing:
– Using the new materials, mechanical testing is performed. This currently includes thermal aging, tensile testing, fracture toughness, impact testing, and creep-fatigue.
� Materials Performance Criteria and Methodology:– Licensing needs for advanced materials and resolution of design
methodology are key long-term needs for application of advanced materials.
� Environmental Testing (Sodium Compatibility):– Information on sodium compatibility of selected alloys is very
limited. Sodium testing capability was reconstituted at ANL and ORNL and through an FOA at the University of Wisconsin.
Tensile Property Comparison
� TMT increased strength without the significant redu ction in ductility (e.g., Gr.92-1-TMT).
� Cu-alloyed modified-Gr.92 steels showed superior st rength but poor tensile ductility (e.g., 9Cr-2WCuVNb and 9Cr-2WCuVNbTiC).
� The optimized-Gr.92 (9Cr-2WVNbC) and modified-Gr.92 (9Cr-1WVTaC and 9Cr-2WVNbTiC) are downselected for further study in FY12 due to th eir significant enhancement in strength without noticeable trade-of f in tensile ductility.
� A new alloy composition (advanced HT-UPS) was propo sed.– Fe-13Cr-16Ni-Mo-Nb-C-N (wt%) base
– Goal is to improve weldability, while maintaining lower material cost.
– 50lb cast ingot arrived.
– A half of the ingot has been processed (hot-forged and hot-rolled).
� A new alloy composition (advanced HT-UPS) was propo sed.– Fe-13Cr-16Ni-Mo-Nb-C-N (wt%) base
– Goal is to improve weldability, while maintaining lower material cost.
– 50lb cast ingot arrived.
– A half of the ingot has been processed (hot-forged and hot-rolled).
Preparation of Advanced HT-UPS
Cast ingot After hot-forging After hot-rolling
� Above goal is achieved with this new alloy composit ion.� Above goal is achieved with this new alloy composit ion.
23
Environmental testing of Advanced Materials
� Environment effects (thermal, radiation and/or cool ant) can have a significant impact on mechanical performance and alloy stability
� Thermal Aging – Time at temperature may degrade material properties.
� Irradiation Testing – Initial irradiation and PIE on candidate alloys will start in
FY09– Initial Testing will help prioritize PIE from MATRIX-II– Some HT-UPS samples from FFTF/MOTA experiments
have also been identified– Data interpretation and semi-empirical modeling will guide
future tests and needs� Corrosion in Sodium
– Corrosion in liquid metals must be evaluated and understood for the candidate alloys
– The pumped-Na loop at ANL will be utilized in addition to convection-driven loops at ORNL
– Initial burden-modeling activities will also provide insight into transfer of C, O, and/or N around the reactor loop
Modeling Creep-Fatigue Behavior of Ferritic Steels
24
� Established an updated database for G91 steel: stress relaxation data were recovered from 13 creep-fatigue tests completed in early 1990s at ORNL.
Cyclic Softening Model
� Developed the Damage Rate Based Model to predict hold-time creep-fatigue life using tensile, creep and fatigue properties
Damage Rate Based Life Prediction Model
� Using the existing database, developed the mechanism-based Cyclic Softening Model and the Stress Relaxation Model by incorporating strain hardening and dislocation creep mechanisms.
Stress relaxation data recorded on strip charts
Digitized stress relaxation data.
Stress Relaxation Model
In-service Inspection Technology Development:Under-Sodium Viewing
25
USV System Evaluation −Sodium Tests
26
Sodium TestsScanning Size: 1” x 1”; Resolution: 50 pixel/inch
Lateral Resolution: 1mm; Vertical Resolution 0.5mm
Test Tank
Dump Tank
Scanning System
Temperature Control Modules
Gold-plated Focal Lens
Transducer
Waveguide
Target 1” away
Target 1.25” away
Advantages of Argonne UWT:
� Reduce the background noise and spurious signal
� Minimize the waveguide attenuation
� Create a clear window to detect target reflection
� Maximize the signal to noise ratio
Argonne WGT Prototype (D=0.625”, L=12”)
Intensity Image
Time-of-Flight Image 12” UWT @650OF
18” UWT @310OF
1st Generation USV Test FacilityIn
tensi
ty,
V
Time-of-Flight, µs
USV Ongoing and Future Works
27
� Linear UWT Array• Fabricate and evaluate linear UWT array• Develop signal and image processing software• Demonstrate crack detection/loose-part identification
capabilities
� Develop brush-type linear UWT Array• Construct a brush-type waveguide for side-view
imaging application• Develop dry-coupling technique for linear UWT array
application• Develop the imaging algorithm applicable to the linear
UWT array
� Sodium Test• Complete new USV test facility for in-sodium tests of
linear UWT array (ANL) and phased array transducer (PNNL)
• Demonstrate crack detection/loose-part identification capabilities in sodium
� US-Japan Collaboration
2nd Generation of USV Test Facility
Dump Tank
Cold Trap
Sodium
PumpFlowmete
r
Positioning System
Test Tank
UWT Array
Positioning System
Advantages:
� Larger opening and tank for linear array
� Larger Testing Components/Targets
� Easier target and transducer setup and replacement
� Add an EM pump (5G/min) and cold trap to circulate and clean sodium
� Better temperature control and Higher Operation Temperature (650OC)
� Closed-loop design for Safer operation
DAQ & Image
Processing
Sodium Flow Control Center
Heating Control Center
Pulser Receiver
Gain Filter
Gas Flow Control Center
UWT Linear Array
Brush-type Linear UWT
ArrayTime-of-Flight Intensity
High temperature linear ultrasonic array
Acoustic field map exhibits significant
inter-element crosstalk and inconsistent
coupling between piezo-elements and
nickel alloy faceplate
• Robust array for immersion (Mk 1)
• 12 discrete transducer elements
• Lead meta-niobate piezoelectric
• Frequency = 2MHz
• Faceplate polished nickel foil
• Operating in liquid sodium (260C)
• Demonstrating array imaging in sodium
Ultrasonic Image of Joyo Fuel Subassembly Mockup
Loose pin
C-scan image
A-scan image
Top of MARICO pin
Top of fuel pins
Bottom plate
Loose MARICO pin
Advanced Energy Conversion Technology
30
Supercritical CO 2 Cycle Coupled to an Advanced Liquid Metal Reactor
Page 31
Turbine Generator
Compressor/Recompressorwith turbine
Recuperators (8)CO2 Storage
Advanced BOP Trade Study – Summary and Conclusion
� Pre-conceptual design of 1,000 MWe supercritical CO2 cycle BOP was developed by vendor - PWR
� Design contained sufficient detail for cost & schedule estimates to ± 35%� BOP Sizing studies were completed – 1 x 1,000 MWe configuration was
selected for this study� Piping, Heat Exchanger, Turbomachinery designs were completed� Preliminary cost comparison (2011$) between steam and S-CO2 plant
completed� The turbine building for S-CO2 plant about 50% the size of steam plant,
however not much floor area for maintenance or equipment laydown
� Results are based upon an Interim Report – it will be revised as appropriate once we receive responses to our comments from the vendor team
32
S-CO2 Brayton Cycle
• Major system upgrades are nearly completed to attain the original testing capabilities of:
– 75,000 RPM on both Turbo-Alternator-Compressor (TAC) units.
– 780 kWth input power.– Peak operating temperature of 1000°F.– Peak operating pressure of 18.7 MPa.– 5.7 kg/sec mass flow rate.– Net power generation on the order of 250 kWe.
• Recent focus has been to:– Complete upgrades.– Perform acceptance testing of the upgraded test
article (TA) at the contractor facility.– Prepare test facility site at Sandia, Albuquerque (see
photos at right).– Transport the TA from the contractor facility to
Sandia Albuquerque.– Perform commissioning tests on the TA to verify
functionality.– Open the Nuclear Technology Users Facility (NTUF)
for commercial component demonstration testing.
Technology Development Efforts
� Contributed to the DOE-EERE Sunshot FOA proposal with NREL, EPRI, and industry participants. This initiative focuses on developing SCO2 Brayton technology for concentrated solar power applications with dry cooling.
� Communicated with industry interests regarding the DOE-NE Brayton cycle TA testing capabilities at NTUF. Likely established NTUF’s first commercial customer.
� Specific objectives for near-, mid-, and long-term testing of the TA are being developed.
� Risks to the smooth and continued access to TA testing are being identified, and contingency plans developed to minimize down time.
Time 3200 s0.916 kg/s T, °C
32.3 P, MPa
152.1 150.7 7.7638.415 8.403 RPM
25,00031.8 19.3
7.702 0.060
140.6 35.078.7 7.764 8.455
8.428
141.6 33.1 35.6 34.6-0.560 0.056 7.754 0.116
1.213 kg/s78.4 35.6
-0.560 -0.560
HeaterAlt
Cooler
CT
LTR
Steady-State Results
� The results of the steady-state calculations from the Plant Dynamics Code (PDC) model are surprisingly close to the experimental data
– Pressures, temperatures, flow rates– Despite all the uncertainties of the
experimental data– Special adjustments for heat loss were
needed
35
Co
de
Pre
dic
tio
n
Experimental Data
0.935 kg/s T, °C32.3 P, MPa
152.9 152.8 7.703 Q, kW8.390 8.383 RPM88.80 25,000
32.3 19.37.707 0.123
3.10145.8 35.1 79.00
80.3 7.751 8.4108.390 -6.55
139.8 35.1 36.4 35.07.735 8.408 7.711 0.116
1.203 kg/s80.4 36.5
8.399 7.719
163.90
PDC Steady-State Results
4.70
1.50
HeaterAlt
Cooler
CT
LTR
A variety of R&D issues for coupling new energy con version to a fast reactor� Updated small-scale sodium Plugging Phenomena Loop has been
– Modified to assure constant temperature in plugging zone� Fundamental Phenomena for IHX
– Concern with thermal shock for compact heat exchangers– Ability to fill and drain also requires development– Small-scale facility designed to look at freeze and thaw issues
� Na-CO2 Interaction– Focused on sodium-to-CO2 heat exchanger conditions– The apparatus has been procured, testing will start in FY12
� Dynamic Modeling of Supercritical CO 2
– Continues ongoing work formerly under Gen IV FY 2010 Work Package, “Energy Conversion – Brayton Cycle Control Analysis”
– Focused upon S-CO2 cycles for SFR, LFR, and VHTR and system level plant dynamic analyses for S-CO2 cycles
Heat Exchanger – FRTesting and Modeling
1,800 cubic feet per minute (cfm) air blower with variable frequency drive and chiller
Test section and heaters inside of stainless steel air duct half-wall
Installation of test section air cooling system
Fundamental Sodium -CO2Interaction Tests
� Small-scale sodium facility is being assembled to provide fundamental data on interactions between sodium and CO2 released into sodium through stainless steel micro-leak configurations and self-plugging of stainless steel micro-leak configurations under realistic conditions of sodium-to-CO 2 heat exchanger failure
� Envisioned failure mode for compact diffusion-bonded heat exchangers involves formation of microcrackslimiting flow of CO 2 into sodium with possibility of self-plugging of crack channels due to formation of solid reaction products
Advanced Modeling and Simulation
39
Goals, Strategy and Accomplishments
� Goal: Apply modern, high-performance computing tech niques to nuclear reactor modeling
– Use advanced simulation tools to improve safety, reduce cost, explore advanced designs – Provide local data needed to enable predictive fuel performance simulations– Understand and reduce uncertainty of computational models
� Strategy: Develop flexible, mission-agile toolbox f or construction of virtual reactor models
– Adopt multi-scale strategy to enable application to problems relevant to industry using a wide range of computing platforms
– Utilize modular architecture to enable component-wise use by most advanced users or integrated user interface driven application by less advanced users.
– Develop collaborations with customers to define near term applications/demonstrations
� FY11 Accomplishments: Delivered initial modules and continued to demonstrate applicability to SFR simulations
– Modules scheduled for FY11 delivery – Nek5000 v4 (CFD) , MOAB v4 (framework), MC2-3 (cross-sections), and SN2ND (neutron transport)
40
PROTEUS Neutronics Modules
� UNIC transport solver modules– MOC-FE (3-D & 2-D MOC), SN2ND (2nd Order Discrete
Ordinates), PN2ND (2nd Order Spherical Harmonics), NODAL (nodal transport)
• All codes use generalized second order finite element mesh to provide true representation of exact geometry
• Provides variety of homogenization options ranging from conventional nodal methods to fully explicit geometry
� FastRANGE transport solver modules� modernized version of legacy SFR toolset (DIF3D, DIF3D-K, REBUS,
VARI3D, ETOE, FTU, MC2-2, RIGEL, MERMC, TWODANT-ANL) to support coupled multi-physics simulations
� MC2-3 module provides high resolution cross-section libraries for fast spectrum applications
� Codes are developed in sub-version repositories for version tracking and verification is completed by automated buildbot
� Validation efforts in progress using legacy data– Simulations of ZPR-7 experiments completed using SN2ND and MC2-
3 (230 group L5T5 with P3 scattering kernel)– Results shown are for fission in the EU foils and capture in the DU
foils for a selected ZPR-7 loading– Results using SN2ND + MC2-3 were equivalent in accuracy MCNP– Legacy VARIANT code could not obtain a similar solution
41
SHARP Thermal Fluids Modules
� Nek5000 DNS/LES module provides accurate, highly-scalable, high-order spectral element CFD
� Nek5000 URANS module provides simulations of unsteady turbulent flow at lower cost than DNS/LES
� STAR-CCM+ module provides access to steady and unsteady RANS solvers of commercial CFD code STAR-CCM+
� SHARP-IF module uses intermediate fidelity simulation toolset using momentum sources to mimic effects of geometric details
� SAS11 modules– lumped parameter representation of T/H and Structural
Mechanics applicable to full system– provides continued access to legacy SFR fuel
performance models� Codes are developed in sub-version repositories for version
tracking and verification is completed by automated buildbot� Validation efforts in progress with legacy data and
international benchmark data– Most recently completed OECD T-Junction Thermal
Mixing Benchmark– Nek5000 provided best predictions of temperature
measurements among all submissions
Wire-wrap vs. Spacer Grid Study
217-pin Wire-Wrapped Assembly
T-junction Thermal Mixing
SHARP Framework Modules
� MOAB module provides highly scalable data management for mesh based simulations
– Currently integrated into UNIC, Nek5000, Star-CCM+ and DIABLO
� MB Coupler module provides scalable parallel solution transfer between meshes of different types
� MeshKit modules provide efficient, parallel meshing capability for reactor geometries
– MeshKit Generation Library provides consistent API access mesh generation functionalities, including the RGG reactor geometry/meshing tool
– CGM Geometry Library provides support for CAD and other geometry types
• Includes interface to Open.CASCADE, an open-source library for geometry
• compatible with (and can import models from) CUBIT's CGM
� Lasso relations library allows associate of mesh to geometry without requiring software dependency between mesh and geometry libraries
� Meshing capabilities have been successfully demonstrated for a variety of reactor configurations with Nek5000 (open source CFD code) and STAR-CCM+(black box commercial CFD code)
43
Awareness of geometry in addition to mesh simplifies specification of material regions. Efficient solution transferenables use of appropriate mesh type for each physics andeach resolution scale.
Flexible reactor geometry and mesh generation tool supportsboth rectangular and hexagonal lattices, as well as details likewire wrap spacers, assembly cans or bypass flow gaps.
Nuclear Data
44
Nuclear Data Effort onFour Major Topics
February 3, 2011 45Advanced Reactor Concepts Working Group Oak Ridge National
Laboratory
Uncertainty ApplicationsUncertainty Applications
Covariance EvaluationsCovariance Evaluations MeasurementsMeasurements
Advanced Measurement Development
Advanced Measurement Development
Pino PalmiottiINL
Won Sik YangANL
BNL
Michael Herman
BNL
Patrick TalouLANL
Fredrik TovessonLANL
Mike HeffnerLLNL
Tony HillINL
Fredrik TovessonLANL
Mike HeffnerLLNL
Tony HillINL
David AsnerPNNL
Wor
k P
acka
ge M
anag
ers
ARC FCRD
FCRD & ARC
FCRD & ARC
A Time Projection Chamber (TPC) was developed under a NERI-c grant to provide precision fission cross section measurements
� Sub-percent fission measurements will significantly reduce uncertainties that impact reactor and fuel cycle construction, operation and safety costs
� TPC will provide 3D “pictures” of the charged particle trajectories
– Alpha backgrounds removed– Sample auto-radiograph (α particles)– Beam non-uniformities– Multi-actinide targets
� TPC will use thin backing foils (<50 µµµµg/cm 2)– Minimize beam interaction backgrounds– Maximize efficiency– Minimize multiple scattering of fragments– H2 drift gas will also minimize scattering
� TPC will provide data on both fission fragments simultaneously
– Random backgrounds removed (vertex requirement)– Fission vertex with <100 µm resolution (fission
radiograph)
AlphasFission fragments
Co-Funded by NE (FCRD and ARC) and NNSA
Nuclear Data Accomplishments
� Fission data was collected using the nominal fission ratio detectors for U-234 in November and December 2011. Once analyzed this data will complete the uranium series, with cross sections for U-233, 234, 235, 236 and 238.
� A high precision U-235 capture cross section data set will be delivered to evaluators in FY2012. This work was made possible by a new fission tagger (PPAC) developed to improve the accuracy of capture measurements on fissile isotopes.
� The fission TPC successfully collected fission ratios of U-238 to U-235 in the neutron beam at LANSCE in Dec 2011-Jan 2012.
� Plutonium-239 was successfully loaded into the TPC in February 2012, and the instrument installed on the flight path at LANSCE for beam experiments, marking a world’s first in imaging fission from Pu-239.
47
Partially instrumented TPC in the LANSCE flight path
Fast Reactor Fuels R&D
48
AFC-2D Experiment Status
� Inserted September 2008� Burnup objective 40 at.%� 14 cycles of irradiation completed� 685 EFPD through EOC 150B
� Discharged from ATR in December 2011� Cooling in ATR Canal, awaiting shipment to HFEF
East Flux Trap Irradiation Housing
AFC-2E Experiment Status
� Inserted August 2009� Burnup objective >20 at.%� 11 cycles of irradiation complete� 483 EFPD through EOC 150B
� Discharged from ATR in December 2011� Cooling in ATR Canal, awaiting shipment to HFEF
East Flux Trap Irradiation Housing
4
2
5
1
3
67
AFC-3A,3B Began Irradiation in Outboard-A Positions (Cycle 150B, October 2011)
� AFC-3 Design– Cd-shrouded– Rodlet design
unchanged– 5 individually
encapsulated rodlets/position
– Freedom to add/remove individual capsules during any outage
A-10 (AFC-3A)(AFC-3B) A-11
AFC-3 Advanced Metallic Fuel Concept for Ultrahigh Burnup — Irradiation Test Plan
AFC-3A AFC-3B
FY Cycle A10-1 A10-2 A10-3 A10-4 A10-5 A11-1 A11-2 A11-3 A11-4 A11-5
2011 150B
U-10Mo
solid
75% SD
U-10Mo
annular
55% SD
materials
U-10Zr
annular
55% SD
U-1Pd-10Zr
U-2Pd-10Zr
solid
75% SD
U-4Pd-10Zr
solid
55% SD
3 years
U-4Pd-10Zr
annular
55% SD
3 years
materials
U-10Mo
solid
55% SD
4-5 years
U-10Mo
solid
55% SD
6-8 years
2012
151A U-aIn-10Mo
U-bIn-10Mo
annular
75% SD
materials
151B
U-In-Mo
annular
55% SD
3 years
152A
153A
U-10Zr
annular
65% SD
U-10Mo
annular
65% SD
U-10Mo
annular
75% SD
U-In-Mo
U-Pd-Mo
solid
65% SD
2-3 year
U-PuGa-1Pd-Zr
U-PuGa-2Pd-Zr
solid
75% SD2013
153B
154A
154B
155B
U-10Zr
annular
75% SD
U-Pu-MA-Mo
annular
65% SD
U-Pu-MA-Mo
annular
55% SD
3 year
U-Pu-Pd-M-Zr
solid
75% SD2014
156A
156B
157A
157B
SD = smear density
Initial "Challenge" problems defined
Initial 1st year experiments defined
2nd and 3rd year experiments proposed
Materials experiments
Experiment discussed, but not finalized
Short-term, early PIE to
demonstrate feasibility
Long-term to demonstrate the
““““Grand Challenge””””
� Fast reactor R&D is focused on key technologies inn ovations for performance improvement (cost reduction) and safety1. System Integration and Concept Development2. Safety Technology3. Advanced Materials4. Ultrasonic Viewing5. Advanced Energy Conversion (Supercritical CO2 Brayton cycle)6. Reactor Simulation7. Nuclear Data8. Advanced Fuels
� Fast reactors have flexible capability for actinide management– A wide variety of fuel cycle options are being considered
� International R&D collaboration pursued in Generati on-IV and multi-lateral arrangements
Summary
53
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