non-lwr scale activities

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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

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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

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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

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SCALE is integral part of NRC IAP Strategy 2, Volumes 3+5

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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.

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Volume 3 focuses on Severe Accident

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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

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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

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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

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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

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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

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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.

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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

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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

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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

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INL A Control Drum Rotation Flux Animations

Shutdown rods inShutdown rods out

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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)

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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

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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.