Post on 26-Oct-2021
transcript
ORNL is managed by UT-Battelle, LLC for the US Department of Energy
Non-LWR SCALE Activities2021 SCALE Users' Group Workshop
Presenter: W. Wieselquist
Contributors:J.W. BaeB. BetzlerF. BostelmannA. LoR. KileG. IlasK.L. ReedA. ShawS. SkutnikE. Walker
22 2021 SCALE Users' Group Workshop
Outline• Current Activities
– NRC non-LWR Severe Accident• HTGR• HPR• FHR
– MSR PIRT– Nuclear data gap analysis NUREG– Nuclear data needs workshops (WANDA, WONDRAM)
• Future Activities– NRC non-LWR Severe Accident
• MSR• SFR
– NRC non-LWR Fuel Cycle Safety
33 2021 SCALE Users' Group Workshop
NRC Integrated Action Plan (IAP) for Advanced Reactors
Near-Term Implementation Action Plan
Strategy 1Knowledge, Skills,
and Capacity
Strategy 2Analytical Tools
Strategy 3Flexible Review
Process
Strategy 4Industry Codes and Standards
Strategy 5Technology
Inclusive Issues
Strategy 6Communication
ML17165A069
44 2021 SCALE Users' Group Workshop
SCALE is integral part of NRC IAP Strategy 2, Volumes 3+5
ML20030A177
ML20030A174 ML20030A176
ML20030A178ML21085A484
Introduction Volume 1
Volume 2Volume 3
Volume 4 Volume 5ML21088A047
These Volumes outline the specific analytical tools to enable independent analysis of non-LWRs, “gaps” in code capabilities and data, V&V needs and code development tasks.
55 2021 SCALE Users' Group Workshop
Volume 3 focuses on Severe Accident
66 2021 SCALE Users' Group Workshop
Volume 3 SCALE Activities
• Understand severe accident behavior• Provide insights for regulatory guidance
• Facilitate dialogue on NRC staff’s approach for source term
• Demonstrate use of SCALE and MELCOR• Identify accident characteristics and uncertainties
affecting source term
• Develop publicly available input models for representative designs
Goals• Build MELCOR full-plant input model
– Use SCALE to provide decay heat and core radionuclide inventory
• Scenario selection
• Perform simulations for the selected scenario and debug
– Base case– Sensitivity cases
Approach
By October 1, 2021:Full-plant models for three representative non-LWRs
• Heat pipe reactor – INL Design A• Pebble-bed gas-cooled reactor – PBMR-400• Pebble-bed molten-salt-cooled – UC Berkeley Mark I
By end of project:• Molten-salt-fueled reactor – MSRE• Sodium-cooled fast reactor – To be determined
Project Start: December 2019Project End: April 2022
77 2021 SCALE Users' Group Workshop
Broad LandscapeHigh-Temperature Gas-Cooled Reactors(HTGR)
Liquid Metal Cooled Fast Reactors(LMFR)
Molten Salt Reactors(MSR)
GEH PRISM (VTR)
Advanced Reactor Concepts
Westinghouse
Columbia Basin
Hydromine
Framatome
X-energy *
StarCore
General Atomics
Kairos (Hermes|RTR)
Terrestrial *
Thorcon
Flibe
TerraPower/GEH (Natrium)*
Elysium
Liquid Salt Fueled
TRISO Fuel
Sodium-Cooled
Lead-Cooled
Alpha Tech
Muons
MicroReactors
Oklo
Stationary
Transportable
Ultra Safe |RTR
Radiant |RTR
Westinghouse (eVinci)
Liquid Salt Cooled X-energy
BWX Technologies
Southern (TP MCFR) |RTR
Oklo
ARDP Awardees
MIT
ACU |RTR *
ARC-20
Demo Reactors In Licensing Review
Risk Reduction * Preapplication
RTR Research/Test Reactor
LEGEND
General Atomics (EM2)
Kairos *
TerraPower
Advanced Reactor Designs
88 2021 SCALE Users' Group Workshop
Reactor Archetypes and Strategies• Heat Pipe Reactor (HPR)
– small size, low burnup, no fuel reshuffle– Continuous Energy (CE) Monte Carlo (MC)
with Depletion (TRITON-KENO or TRITON-Shift)
• High-Temperature Gas Reactor (HTGR)– high-burnup, continuous reload– reference: Multigroup (MG) MC TRITON +
ORIGEN iterative equilibrium core inventory– production: ORIGAMI
• Fluoride salt-cooled, High-temperature Reactor (FHR)– high-burnup, continuous reload– reference: Multigroup (MG) MC TRITON +
ORIGEN iterative equilibrium core inventory– production: ORIGAMI
• Molten Salt-fueled Reactor (MSR)– liquid fuel– reference: TRITON-NEWT (MG 2D) with new
new flow input + ORIGEN loop distributions– production: ORIGAMI
• Sodium Fast Reactor (SFR)– high burnup, batch fuel reload– reference: MG MC– production: ORIGAMI
99 2021 SCALE Users' Group Workshop
General ORNL Methodology for Fuel Inventory• ORNL has used a methodology
with the Oak Ridge Isotope GENeration(ORIGEN) code to rapidly generate inventories using ORIGEN reactor libraries
• SCALE/ORIGEN use of fundamental nuclear data allows the following to be calculated from nuclide inventory (moles of each nuclide in a system)
– mass– decay heat– activity– gamma emission– neutron emissions
• With SCALE 6.2 (2016), the sequence ORIGAMI was released which is the modern approach of using ORIGEN reactor libraries
1010 2021 SCALE Users' Group Workshop
Plans for SCALE/ORIGAMI and HTGR
• Soon ORIGAMI will have a new PBMR-400 Fuel Type and the ability to generate (in seconds)– fuel inventory for a
PBMR-400 pebble – initial enrichment– specific power history– cooling time
• Generalizing what we learn for the PBMR-400 will enable future HTGR Fuel Types
>50 different fuel types supported!
Current Fuel Types
1111 2021 SCALE Users' Group Workshop
Aspects of the ORNL methodology for fuel inventory
• Rapid answers to common questions such asWhat I/Cs/Pu content could I expect in a PBMR-400 pebble at 90 GWd/MTU?
a. assuming constant power?b. pass-dependent power?c. during a power maneuver?d. after 4 days of decay?e. after 40 days of decay?f. after 40 years of decay?g. at 80 GWd/MTU?h. in a pebble with +1% enrichment?
• Up-front work required– Sensitivity analysis of the reactor system to
understand the state changes that impact neutron flux spectrum in the fuel (e.g. moderator density in BWR)
– Running many CPU-hours of TRITON coupled transport+depletion cases to generate a database of 1-group cross sections 𝜎𝜎 which can be interpolated to a specific state (ORIGEN reactor library)
– Those libraries can then be used later (in ORIGAMI) to regenerate inventory and reaction rates: 𝑅𝑅𝑅𝑅(𝑡𝑡) = 𝜎𝜎(𝑡𝑡) 𝑁𝑁(𝑡𝑡) 𝜙𝜙(𝑡𝑡)
– Why do it this way? If 𝜎𝜎 is insensitive to decay time, power level, then b through h can be answered from a single TRITON pre-calculation!
Each answer requires a <10 second calc. on a single CPU
Why is speed important? This approach is not just for seeding MELCOR nodalizations. All back-end analysis can use this approach: dry storage casks, on-site storage, discharge inventory analysis, transportation packages.
1212 2021 SCALE Users' Group Workshop
HTGR PBMR-400Lead: Steve Skutnik• Key assumptions
– License applications will specify pebble circulation strategy and equilibrium core
– Analyzing the equilibrium core is the limiting case from an inventory/decay heat standpoint
• Related Work• NGNP provided significant code development and
validation basis for TRISO Fuels
• Recent Accomplishments – TM describing HTGR neutronics characteristics– Journal paper overviewing SCALE methodology– NRC staff & public demo complete
• Current Work– ORIGAMI implementation for pebble systems (early 7.0
betas)
PBMR-400
1313 2021 SCALE Users' Group Workshop
FHR Berkeley Mark 1Lead: Rike Bostelmann
• Key assumptions– License applications will specify pebble circulation
strategy and equilibrium core– Analyzing the equilibrium core is the limiting case
from an inventory/decay heat standpoint
• Related Work• Robby Kile is performing SA/UQ for the benchmark
https://kairospower.com/generic-fhr-core-model/ with SCALE+MELCOR
• Recent Accomplishments – Equilibrium iteration strategy– Delivered decay heat, inventory to MELCOR team
• Current Work– TM and NRC public demo prep in progress
BK MK 1
1414 2021 SCALE Users' Group Workshop
HPR INL Design ALead: Erik Walker
• Key assumptions– Once-through core is fairly straightforward to model
with CE MC – Focus on validation
• Recent Accomplishments – Finalized model & results– NRC public demo
• Current Work– TM in progress– Open source repository for models
INL A
Fuel
Potassium heat pipe
Fuel element latticeControl drum
200 cmcore height
1515 2021 SCALE Users' Group Workshop
INL A Control Drum Rotation Flux Animations
Shutdown rods inShutdown rods out
1616 2021 SCALE Users' Group Workshop
Verification & validation of INL Design A SCALE models• Verification
– Compared to INL A reference design description• Axial power shape• Control drum worth
– Multi-group (faster) vs. continuous energy physics (more accurate) shows an average ~150 pcm higher reactivity
– ENDF/B-VIII.0 vs. ENDF/B-VII.1 shows an average ~300 pcm lower
• Validation– 1% +/- 2% bias in decay heat based on burst-fission experiments
(90% fast fission in U235 during lifetime)– 200 pcm +/- 400 pcm bias in eigenvalue based on 24 critical
experiments with >90% similarity (defined as ck>0.9) to beginning-of-life (BOL) cold zero power (CZP)
1717 2021 SCALE Users' Group Workshop
Summary• SCALE team is performing non-LWR work through at least 2022
• Our focus is on• code readiness for confirmatory analysis• integrated analyses with MELCOR for severe accident and fuel cycle
safety issues• exposing important nuclear data gaps• exposing important validation gaps
• Deliverables for non-LWR severe accident project• ORNL TM reports describing inventory & decay heat calculations• Openly available SCALE model repositories for the 5 prototype non-
LWRs
1818 2021 SCALE Users' Group Workshop
List of References• Sensitivity/uncertainty analysis with TSUNAMI (perturbation theory):
– B. L. Broadhead, B. T. Rearden, C. M. Hopper, J. J. Wagschal, and C. V. Parks (2004), “Sensitivity- and Uncertainty-Based Criticality Safety Validation Techniques,” Nucl. Sci. Eng., 146(3), pp. 340–366.
– B. T. Rearden, M. L. Williams, M. A. Jessee, D. E. Mueller, D. Wiarda, (2011). Sensitivity and uncertainty analysis capabilities and data in SCALE. Nuclear Technology, 174(2):236–288.
• Depletion perturbation theory (DPT):– Keith C. Bledsoe, Germina Ilas, Susan L. Hogle, “Application of Depletion Perturbation Theory for Sensitivity Analysis in the
High Flux Isotope Reactor” Trans. Am. Nucl. Soc., 121 Nov. 2019
• Sensitivity/uncertainty analysis with Sampler (random sampling approach):– B. L. Broadhead, B. T. Rearden, C. M. Hopper, J. J. Wagschal, and C. V. Parks (2004), “Sensitivity- and Uncertainty-Based
Criticality Safety Validation Techniques,” Nucl. Sci. Eng., 146(3), pp. 340–366.– F. Bostelmann (2020), “Systematic Sensitivity and Uncertainty Analysis of Sodium-Cooled Fast Reactor Systems,” École
polytechnique fédérale de Lausanne, Switzerland. https://infoscience.epfl.ch/record/274286 – F. Bostelmann, D. Wiarda, W. Wieselquist (2021), “Extension of SCALE/Samplers’ Sensitivity Analysis,” Annals of Nuclear
Energy, submitted.
• Analysis:– F. Bostelmann, G. Ilas, and W. A. Wieselquist (2020), “Key Nuclear Data Impacting Reactivity in Advanced Reactors,”
ORNL/TM-2020/1557, 2020. https://info.ornl.gov/sites/publications/Files/Pub140896.pdf – F. Bostelmann, G. Ilas, C. Celik, A. Holcomb, W. Wieselquist (2021), “Nuclear Data Performance Assessment for Advanced
Reactors,” ORNL/TM-2021/2002, NUREG, submitted for review to NRC.