ML21179C060

Presentation - Scale/Melcor non-LWR Source Term Demonstration Project - Heat Pipe Reactor

ML21179C060
Person / Time
Site:	Grand Gulf, Arkansas Nuclear, River Bend, Waterford
Issue date:	06/28/2021
From:	Jordan Hoellman NRC/NRR/DANU/UARP, Oak Ridge, Sandia, US Dept of Energy, National Nuclear Security Admin
To:
Hoellman J
References
DE-NA0003525
Download: ML21179C060 (91)

Text

SCALE/MELCOR Non-LWR Source Term Demonstration Project -

Heat Pipe Reactor June 2021 Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energys National Nuclear Security Administration under contract DE-NA0003525. SAND20XX-XXXX P

Outline NRC strategy for non-LWR source term analysis Project scope Heat pipe reactor fission product inventory/decay heat methods and results Heat pipe reactor plant model and source term analysis Summary Appendices

• SCALE overview

• MELCOR overview 2

Integrated Action Plan (IAP) for Advanced Reactors Near-Term Implementation Action Plan Strategy 1 Strategy 4 Knowledge, Skills, Industry Codes and Capacity and Standards Strategy 5 Strategy 2 Technology Analytical Tools Inclusive Issues ML17165A069 Strategy 3 Strategy 6 Flexible Review Communication Process 3

IAP Strategy 2 Volumes 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.

Introduction Volume 1 ML20030A174 ML20030A176 Volume 3 Volume 2 Volume 4 Volume 5 ML20030A177 ML21085A484 ML21088A047 ML20030A178 4

NRC strategy for non-LWR analysis (Volume 3) 5

Role of NRC severe accident codes 6

Project Scope Project objectives Understand severe accident behavior

• Provide insights for regulatory guidance Facilitate dialogue on staffs 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 8

Project scope Full-plant models for three representative non-LWRs (FY21)

• Heat pipe reactor - INL Design A

• Pebble-bed gas-cooled reactor - PBMR-400

• Pebble-bed molten-salt-cooled - UC Berkeley Mark I FY22

• Molten-salt-fueled reactor - MSRE

• Sodium-cooled fast reactor - TBD 9

Project approach

1. Develop SCALE model to provide MELCOR with decay heat, core radionuclide inventories, kinetics parameters, power distribution

2. Build MELCOR full-plant input model

3. Scenario selection

4. Perform simulations for the selected scenario and debug

• Base case

• Sensitivity cases

• Uncertainty cases 10

Advanced Reactor Designs Broad Landscape Liquid Metal Cooled Fast High-Temperature Gas-Cooled Molten Salt Reactors Micro Reactors Reactors (MSR) Reactors (LMFR) (HTGR)

TerraPower/GEH (Natrium)* Kairos

• Westinghouse (eVinci)

X-energy

• GEH PRISM (VTR) Kairos (HermeslRTR) BWX Technologies Framatome Oklo Liquid Salt Cooled X-energy StarCore Advanced Reactor Radiant lRTR Concepts MIT Sodium-Cooled Terrestrial

• Transportable TerraPower Westinghouse TRISO Fuel Ultra Safe lRTR Southern (TP MCFR) lRTR Columbia Basin General Atomics (EM2) ACU lRTR

• Oklo Hydromine General Atomics Elysium Stationary Lead-Cooled Thorcon LEGEND ARDP Awardees Muons Demo In Licensing Review Flibe Reactors Risk Reduction

• Preapplication Alpha Tech ARC-20 RTR Research/Test Reactor Liquid Salt Fueled 11

Heat Pipe Reactor Heat pipe for reactor use Construction

• Metal pipe with wick along pipe inside surface

• Liquid coolant fills area between wick and pipe inside surface Operation

• The core heats the liquid coolant which generates vapor

• The vapor flows to the other end of the heat pipe where it condenses, heating the secondary system fluid

• Coolant film return flow by capillary forces 13

Heat pipe wick being installed 14

First heat pipe reactor KRUSTY experiment

• Kilowatt Reactor Using Stirling TechnologY

• Part of NASAs Kilopower project

• 3 kW thermal power

• 8 heat pipes clamped to uranium cylinder

• Heat pipes transfer heat to Sterling engine

• Operated 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> (March 20-21, 2018)

• LANL video on Kilopower 15

Publicly available designs LANL Megapower

• Megawatt-size heat pipe reactor

• Described in LA-UR-15-28840 INL Designs A and B

• Two alternatives to Megapower for improved performance and ease of construction

• Described in INL-EXT-17-43212, Rev 1 16

LANL Megapower versus INL Design A LANL Megapower

• UO2 High-Assay Low-Enriched Uranium (HALEU) fuel

• Fuel region contained between the top and bottom reflector assemblies

• Negative temperature coefficient from Doppler broadening and axial elongation

• Passive removal of decay heat INL Design A

• To address potential issues with manufacturing and defense in depth No stainless-steel monolith (reduces thermal stress, intended to simplify construction)

The fuel is encased in stainless steel cladding Heat pipes fabricated separately and inserted into central hole in fuel element

• Used for SCALE/MELCOR demonstration project 17

INL Design A (1/2)

Reactor

• 5 MW thermal power

• 1134 fuel elements UO2 HALEU fuel (5.2 MT)

Annular fuel elements with with stainless steel cladding on both sides Outside of fuel element has hexagonal shape HP at the center of each fuel element

• 1134 heat pipes Potassium at 650 to 750 C Vertical orientation for gravity-assisted performance 1.8 cm outside diameter

• 2 emergency control rods of B4C

• 12 alumina control drums with arcs of B4C for reactivity control 18

INL Design A (2/2)

Reactor

• 3 neutron reflectors (top, side, bottom) around the core Top/bottom reflectors are stainless steel +

beryllium oxide (BeO)

Side reflector is alumina (Al2O3)

• Radiation shield surrounds the core 5.08 cm stainless steel core barrel [INL-EXT-17-43212, Rev 1]

15.24 cm B4C neutron shield Reactor Secondary system

• Open-air Brayton cycle Operates at 1.1 MPa 1.47 MW electrical power output (29% efficiency)

[INL/CON-17-41817] 19

Heat Pipe Reactor Fission Product Inventory/Decay Heat Methods and Results

Workflow SCALE specific Generic End-user specific SCALE Inventory Other Binary Output Interface File MACCS Input SCALE Power SCALE Text distributions MELCOR Input Output Kinetics data

• SCALE capabilities used

• Relatively small amount of data except for

• KENO or Shift 3D Monte Carlo transport nuclide inventory

• ENDF/B-VII.1 continuous energy physics

• new interface file developed for inventory using standard JSON format

• ORIGEN for depletion

• easily read in python and post-processed into

• Sequences MELCOR or MACCS input CSAS for reactivity (e.g. rod worth)

• contains nuclear data such as decay Q-value TRITON for reactor physics & depletion for traceability when performing UQ studies 21

INL Design A Neutronics Summary

• 5 MWt rated power with 5-year

• 2.0 GWD/MTU discharge burnup operating lifetime

• 1,134 heat pipe/fuel element units

• UO2 fuel with 19.75% 235U enrichment

• Discretized with 20 axial and 5 radial fuel zones

• 4.57 MTU initial core loading

• 1.0951 MW/MTU specific power

~100 cm 200 cm core height Potassium Helium Alumina heat pipe void reflector Fuel Control drum Fuel Element Lattice Cross Section at Midline 3D Core Model 22

HPR-Related SCALE Updates

• New fast-spectrum nuclear data library

• New 302-group structure was developed based on group structures optimized for fast systems (sodium-cooled fast reactors)

• Enables fast depletion

~6.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ~1.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> using KENO

• Added 3D data visualization

• Input geometry

• Mesh data overlay (flux, fission source)

• Probability table update for unresolved resonance region for fast systems

• ~400 pcm error for fast-reactor systems with Pu

• Shift integration

• Full-core continuous-energy (CE) depletion is tractable for HPRs and TRITON-Shift scales to 10,000s of cores for faster turnaround (TRITON-KENO only scales to <100s of cores)

• Developed MADRE test suite for advanced reactors

• Finds equivalent multigroup (MG) vs. CE performance

• ENDF/B-VIII.0 ~400 pcm less than ENDF/B-VII.1 3D SCALE model of INL Design A 23

HPR-Related SCALE Updates

• Example of 3D beginning-of-life (BOL) flux map overlay Normalized flux Front (X-Z) Top (X-Y) 3D 24

Modeling Assumptions

• Full-core 3D Monte Carlo with continuous energy physics

• System state defined in INL report

• Temperature Fuel iteration 1: uniform 1,000K iteration 2: informed by MELCOR temperature profile Working fluids 950K Reflector 950K

• Geometry Annular fuel Thermal expansion of fuel stack (UO2) radial reflector (Alumina) fuel cladding (stainless steel [SS])

3D Fission Rate 25

Verification and Validation

• Verification

• Compared to INL A reference design description Axial power shape Control drum worth

• Multigroup (faster) vs. continuous energy physics (more accurate) comparison shows an average ~150 pcm higher reactivity Years 0.00 0.42 0.84 1.25 1.67 2.08 2.50 2.92 3.33 3.75 4.17 4.58 5.00 MG-CE Diff (pcm) 141 60 62 260 122 145 104 134 198 149 305 112 128

• ENDF/B-VIII vs. ENDF/B-VII.1 comparison shows an average ~300 pcm lower Years 0.00 0.01 0.42 0.84 1.25 1.67 2.08 2.50 2.92 3.33 3.75 4.17 4.58 5.00 VIII.0-VII.1 Diff (pcm) -272 -302 -288 -295 -315 -288 -298 -327 -306 -309 -259 -288 -280 -333

• Validation basis

• 1% +/- 2% bias in decay heat based on burst-fission experiments

• 200 pcm +/- 400 pcm bias in eigenvalue based on 24 critical experiments with similarity index ck>0.9 compared to BOL cold zero power 26

Unit Cell Verification

• INL reported sensitivity study results of k-inf on pitch and clad thickness changes using infinite unit cell models in MCNP

• These models were replicated in SCALE

• Using identical explicit isotopics with ENDF/B-VII.0 library used in the INL report

• Using the SCALE standard composition library with ENDF/B-VII.1 library

• SCALE k-inf results differed by roughly 50 pcm with identical models

• Updating library and material definitions added, on average, 50 pcm to the SCALE results Case Outer SS Pitch (cm) kinf1 (MCNP) kinf (SCALE) kinf (SCALE)

Clad (cm) ENDF 7.0, Isotopics ENDF 7.1, Std Comp 1 0.1 2.786 1.25953 1.259937 (40.7) 1.260461 (93.1) 2 0.05 2.786 1.27496 1.275447 (48.7) 1.275798 (83.8) 3 0.05 2.686 1.28830 1.288864 (56.4) 1.289494 (119.4) Design A unit cell with

1. Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor, NL/EXT-17-43212, May 2018 reflective boundary conditions1 27

Full Core Verification

• Reactivity control devices were tested in different configurations

• All Poisons Out - Both shutdown rods were withdrawn, and control drums (CDs) were turned away

• All Poisons In - Both shutdown rods were inserted, and CDs were turned in

• Comparisons were done using both the explicit isotopics with ENDF/B-VII.0 library and standard composition library with ENDF/B-VII.1

• Reactivity worth calculations were performed and compared to reference results

• Identical models agree well with < 200 pcm k-eff differences and < 3.5% reactivity worth differences Control Condition/Parameter Design A MCNP Design A SCALE Design A SCALE ENDF 7.0, Isotopics ENDF 7.1, Std Comp All Poisons Out 1.02825 1.029816 (156.6) 1.02989 (164.0)

All Poisons In 0.84594 0.846039 (9.9) 0.84526 (-68.5)

Control Drums In 0.95042 0.95067 (25.0) 0.950304 (-11.6)

Annular Shutdown Rod In 0.94555 0.947445 (189.5) 0.94725 (170.0)

Solid Shutdown Rod In 0.95933 0.960734 (140.4) 0.960660 (133.0)

=0.007 =0.0072 (20) =0.0072 (20)

BOL Excess Reactivity ($) 3.925 4.021 (2.5%) 4.025 (2.6%)

Total Drum Worth ($) 11.377 11.228 (-1.3%) 11.278 (-0.9%)

Individual Drum Worth ($) 0.970 0.985 (1.5%) 0.990 (2.1%)

Annular Shutdown Rod Worth ($) 12.151 11.725 (-3.5%) 11.749 (-3.3%)

Solid Shutdown Rod Worth ($) 9.981 9.698 (-2.8%) 9.705 (-2.8%) Effect of control drum rotation on eigenvalue 28

Control Drum Rotation Flux Animations Shutdown rods out Shutdown rods in 29

Validation Basis: Short-Term Decay Heat

• Fissioning nuclides in INL A

• 90% from 235U

• 10% from 238U Differential Energy Release

• Negligible from Pu

• Cumulative energy release following shutdown

• ~90% by 0.3 days

• ~92% by 1 day

• ~96% by 10 days (MeV/fission)

• Burst fission experiments measure energy release over time (t<1 day) from a single fission of 235U

• Most accurate measurements in the set have 1-sigma uncertainty in the 2-3%

range

• ORIGEN simulation is within 2-sigma uncertainty bounds shown in figure for almost all measurements

• Based on burst-fission data analyzed so far, 1% +/- 2% bias in instantaneous decay heat recommended for SCALE modeling of INL A 30

Decay Heat after Core Shutdown Decay heat for select isotopes for the INL Design A at 2 GWd/MTU 1.00E+05

• Top 10 decay heat producing isotopes 1.00E+04 in the first 10 days la140 following shutdown Decay Heat (watts/MTIHM) pr144 np239 i132

• Subtotal shows 1.00E+03 nb95 sum of top 10 zr95 y91 sr89 ba140 ru103 Subtotals 1.00E+02 Totals 1.00E+01 0 1 2 3 4 5 6 7 8 9 10 Time (days) 31

Why Is Decay Heat So Much Lower than a Pressurized Water Reactor (PWR)?

1.0E+07

• INL A has 2.7% specific power of PWR

• Comparing INL A fuel vs.

PWR fuel per MTIHM 1.0E+06

• PWR at 2 GWd/MTU Decay Heat (W/MTIHM)

• INL A has 2.9% of PWR decay heat at t=0 1.0E+05

• INL A has 4.8% of PWR decay heat at t=10 days

• PWR at 60 GWd/MTU

• INL A has 3.1% of PWR 1.0E+04 decay heat at t=0

• INL A has 2.6% of PWR decay heat at t=10 days

• Does this mean decay 1.0E+03 0 1 2 3 4 5 6 7 8 9 10 heat can be scaled with Time (days) specific power?

INL A 2 GWd/MTIHM PWR 2 GWd/MTIHM PWR 60 GWd/MTIHM 32

Scaling HPR Decay Heat Curve from a PWR 1.0E+05 15.0%

• Comparing INL Design A with PWR decay 10.0% heat curve scaled down Decay Heat (watts/MTIHM) 5.0%

• INL A differs by 13.2% at t=0 Difference (%)

-10.0% at t=1.88 1.0E+04 0.0%

-5.2% at t=10

-5.0%

• Decay heat does not scale using specific

-10.0%

power 1.0E+03 -15.0%

0 1 2 3 4 5 6 7 8 9 10 Time (days)

INL A 2 GWd/MTIHM PWR 60 GWd/MTIHM INL A / PWR scaled -1 (%) 33

Activity after Core Shutdown Activity for selected Cs and I isotopes for the INL Design A at 2 GWd/MTU 1.00E+05 1.00E+04 1.00E+03 Activity (Ci/MTIHM) i131 1.00E+02 i132 i133 1.00E+01 cs137 i135 i134 1.00E+00 cs138 1.00E-01 1.00E-02 0 2 4 6 8 10 Time (days) 34

Power Distribution 160

• Axial-normalized power peaking factors LANL Ref agree well with distribution from LANL INL Ref and INL documents Axial Height above Bottom of Active Fuel (cm) 140 0.008 years ORNL

• INL reference gives hottest pin power profile while the LANL and ORNL are core averages 5.000 years ORNL 120

• Peaks at the top and bottom are due to axial reflection

• Not as much reflection in the top due to 100 heat pipes

• MCNP models did not use fine enough mesh to capture bottom reflector peak 80 fully

• Axial peaking does not fluctuate over 60 core lifetime due to low burnup 1.5 Normalized Peaking Factor 1.4 0.008 years ORNL 40 1.3 5.000 years ORNL 1.2 1.1 1 20 0.9 0.8 0.7 0

0.6 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3 0 10 20 30 40 50 Radial Distance from Center (cm) Normalized Peaking Factor 35

Reactivity Feedback Effects

• Four negative reactivity feedback effects reported

• Doppler broadening (primary)

• Fuel axial thermal expansion

• Alumina reflector radial thermal expansion

• Outer clad radial thermal expansion

• Modeled all radial effects Axial Fuel Expansion simultaneously

• Outer SS clad radial expansion Radial Clad Expansion

• Gap closure and increased pitch Feedback Effect (cents/°C) INL ORNL

• Alumina radial expansion Doppler -0.1074 -0.1113 UO2 Fuel Axial Elongation -0.0422 -0.0437

• Control drum drift Alumina Reflector Radial Thermal Expansion -0.0225 -0.0284 Outer SS Fuel Clad Thermal Expansion -0.0323 -

All Radial Expansions (Clad, Reflector, and CDs) - -0.0636 Total -0.2044 -0.2185 36

Neutronics Summary

• Eigenvalue bias was assessed for BOL cold zero power

• 200 pcm +/- 400 pcm based on 24 critical experiments with similarity ck>0.9

• Decay heat bias 1% +/- 2% based on burst-fission measurements

• New 302-group structure was developed

• Demonstrates a ~150 pcm bias over core lifetime compared to CE

• Axial refinement study shows higher reflector peaks at top and bottom of core compared with reference documents

• Using SCALE gives a more realistic representation of decay power than scaling PWR decay power

• 13.2% at shutdown

• -5.2% after 10 days 37

MELCOR Heat Pipe Reactor Model

MELCOR Heat Pipe Reactor Modeling: 1 When present, HPs replace conventional convective heat transfer between the fuel and coolant channel with the energy transfer from the fuel to the evaporative region of the HP.

HP models are special components within the COR package.

Heat rejection from the HP model at the condensation interface is transferred to the CVH package.

Basic geometry of a heat pipe is assumed to be a circular cylinder characterized by a relatively small set of geometric values, e.g.:

• RO outside radius of heat pipe wall (m),

• RI inside radius of heat pipe wall (m),

• Dwick thickness (or depth) of the wick (m), and

• wick porosity of the wick (-).

Axial lengths of the condenser, adiabatic, and evaporator sections are implicitly defined by the COR package cells that these regions are associated with.

39

MELCOR Heat Pipe Reactor Modeling: 2 HP modeling approaches within MELCOR reflect the purpose and constraints of the systems-level integrated code that it is.

MELCOR accommodates HP models of different fidelity through a common interface and a specified wall and working fluid region nodalization.

• Model 1: working fluid region modeled as high thermal conductivity material.

• Model 2: thermodynamic equilibrium of working fluid (sodium or potassium EOS). P, T and liquid/vapor fraction evolve in time.

Sonic, capillary and boiling limits enforced.

• Accepts experimental or design-specific performance limit curves

• Flexible implementation allows for multiple HP definitions in the same MELCOR input deck and multiple HP regions Time-dependent conservation-of-energy equations are solved within the HP component and include boundary conditions linking them with the neighboring fuel (evaporator region) and coolant (condenser region)

Illustrative MELCOR HP component nodalization to define MELCOR variables. Actual nodalization has more nodes.

40

MELCOR Heat Pipe Reactor Modeling: 3

• Core region modeled as a 2-D multi-ring representation Ring 1 - Control Rod inner baffle plate outer baffle plate lower support plate fission gas plenum region BeO reflector region Horizontal cut through core region lower head

• Ring-to-ring radiative exchange implemented through the generalized core heat transfer pathway modeling in MELCOR Vertical cut through core region 41

HP limits of operation Steady-state operational limits modeled in MELCOR

• Sonic limit

• Choked flow of vapor through the central core

• Capillary flow limit

• liquid flow rate at maximum capillary pressure difference

• Boiling limits

• As heat flux increases, both nucleate and film boiling related issues can disrupt heat transfer.

Film boiling can lead to a sudden drop in heat transfer efficiency.

• Condenser HX limit

• The heat exchanger absorbing heat from the condenser may have operation limits of its own.

Each of these limits depend on the HP details (geometry, materials, type of wick, working fluid etc.)

and can independently vary in magnitude based on operational conditions.

Estimated HP SS limits for a design considered in a 2018 INL report INL/EXT-17-43212 42

HP Failure Modes Modeled in MELCOR Several failure modes considered

• HP wall or end-cap failure due to time-at-temperature if the HP is subjected to high operating temperatures and associated pressures, such as might occur in a complete loss of heat sink (e.g., the HX fails)

• Local melt-through of the HP wall due to a sudden influx of heat

• HP wall or micro-imperfections in end-cap welds or HP wall materials after being subjected to time-at-operating temperatures and pressures.

43

MELCOR HP failure modeling INL Design A HP Pressure versus Working Fluid Temperature 10 9

• HP temperature excursion leads to 8

7 working fluid pressurization and HP wall Pressure (MPa) 6 creep failure 5 Larson-Miller model used for wall failure 4

o 3

o Subsequent response includes HP failure and 2 depressurization 1

• Alternate user-specified criteria for HP 0

0 200 400 600 800 1000 1200 1400 1600 Working Fluid Temperature (°C) wall failure o HP wall failure can be a specified event (e.g., initiating event) or as an additional failure following a creep rupture failure (i.e., creep failure is predicted before wall melting)

• Optional user features to dynamically control or disable HP evaporator or condenser wall heat transfer and to start the fuel cell radionuclide leakage 44

MELCOR cascading HP failure modeling HPR model can be subdivided Zone 4: 18 HP elements into an arbitrary number of rings Zone 3: 12 HP elements Zone 2: 6 HP elements

• Generalized input matrix for fuel Zone 1: 1 HP element element connectivity governs heat flows

• Cascade region shown with 4 zones but could be larger

• User specifies HP failure(s)

• Bulk HP response modeled outside of the cascade region

• Consequences of initial failure(s) on adjacent ring responses Zone modeling approach Multiple fuel rod used in SFP applications components in the center assembly and for cascading fuel assembly four peripheral assemblies ignition [NUREG/CR-7216]

45

Heat Pipe Reactor Plant Model and Source Term Analysis

MELCOR model of INL Design A - Reactor Reactor modeling

• 2-D reactor nodalization Ring 1 - Control Rod 14 axial levels Condenser (secondary heat 15 radial rings exchanger)

Level 14

• 14 concentric rings of heat pipes alumina core neutron reflector barrel shield (width of ~1 fuel assembly)

Level 13

• Center ring models the emergency control rod guides inner baffle plate outer baffle plate

• Top and bottom reflectors are in Evaporator axial levels 1 and 13 (fuel elements)

Levels 3-12

• Heat pipes transfer heat to the lower support plate secondary Brayton air cycle in Lower reflector fission gas plenum region BeO reflector region axial level 14 Levels 1-2 lower head

• Core region is surrounded by stainless steel shroud, alumina Ring 1 is the Rings 2-15 are the active core Reflector and neutron shield reflector, core barrel, and B4C control rod guide (each ring = pitch of 1 fuel element) neutron shield 47

Reactor vessel - release pathways Release from fuel to reactor vessel

• Stainless-steel cladding failure at 1650 K Release from reactor vessel to reactor building

• Assumed reactor vessel leakage Heat-pipe release path

• Requires heat-pipe wall failure in two places o Creep rupture followed by melting

• Creep rupture failure in the heat-pipe condenser region (secondary system region) could lead to reactor building bypass 48

Enclosure building nodalization LANL and INL HPR descriptions did not address the enclosure CV1000 building (Environment)

Modeling includes internal building FL5010 (Upper Leakage) circulation flow paths

• Natural circulation into and out of the FL5025 (Lower Leakage) CV5010 (Reactor Building Floor 2) reactor cavity FL5015 FL5020

• Natural circulation within the building CV5005 (Reactor Building Floor 1)

Building leakage addressed FL5000 (Reactor Cavity Flow)

Natural Convection FL5005 (Reactor Cavity Flow)

Natural Convection UP parametrically UP

• Base leakage similar to the reactor CV5000 building surrounding the BWR Mark I (Reactor Cavity) containment Ground 49

MELCOR input model attributes Radionuclide inventory and decay heat at start of accident predicted with SCALE

• End-of-cycle inventory at 5-yr Point kinetics model for transient power calculation Heat transfer between adjacent fuel elements modeled using radiative exchange

• Heat transfer efficiency is parametrically varied Potential for heat pipe creep rupture monitored in the evaporator and the condenser regions Heat pipe limits estimated using LANL HTPIPE code

• MELCOR accepts sonic, capillary, entrainment and boiling limit curves

• Potassium and sodium limit curves were developed

• MELCOR can also accept proprietary performance curves when available 50

Description of the TOP scenario Transient Overpower (TOP) scenario selected for demonstration calculations

• Control drums malfunction and spuriously rotate outward Modeled as linear reactivity insertion rate in $/second

• Safety control rods assumed to insert when peak fuel temperature exceeds 2200 K

• Strong feedback coefficient creates linear power increase Performed sensitivity analysis to show how MELCOR could be used to gain insight into key source term drivers

• Sensitivities focused on source term and HPR parameters

• Previous LWR parameters do not necessarily translate to HPR uncertainties 51

Transient Overpower (TOP) scenario timeline t = -5000 s t=0s Tmax = 2200 K t = 24 h Steady-State Reactivity Insertion Post-SCRAM

• Initialization

• Power increase

• Radial cooling by

• Fuel temperature

• Temperature rise natural processes stabilizes

• Heat pipe failure

• Fission product

• Core damage release and

• Fission product transport release

Transient Overpower (TOP) base scenario (1/7)

Reactor Power 9

The control drums start rotating at t=0 sec, which HPs hit the boiling limit 8

leads to an increase in the core power over 0.9 hr 7

• Negative fuel temperature reactivity feedback limits 6 the rate of power increase Control rods are inserted Power (MW) 5 4

3 The core steadily heats until the maximum heat flux 2 location reaches the boiling limit 1

• The heat transfer rate is limited above the boiling limit, 0 0 1 2 3 4 5 6 which leads to a rapid heatup rate Time (hr)

Maximum Fuel Temperature

• The SS cladding is assumed to fail at 1650 K (just 2400 below its melting point), which starts the fission 2200 Assumed manual SCRAM &

product releases into the reactor secondary isolated 2000

• The reactor is assumed to trip at 2200 K Temperature (K) 1800 1600 Passive radial heat dissipation and heat loss to the reactor cavity Radial heat dissipation and heat loss to the reactor 1400 cavity passively cools the core 1200 Limiting HP location hits the boiling limit

• No active heat removal (secondary system trips and 1000 isolates) 0 6 12 Time (s) 18 24 53

Transient Overpower (TOP) base scenario (2/7)

HP performance limit curves with the TOP response The HP performance limits at the 100000 highest heat flux location show a steady heatup to the boiling limit <1 min heatup to HP & clad failure

• Once the boiling limit is reached, there 10000 is a rapid heatup over the next minute Power (W/HP)

The fuel rapidly heats to melting conditions Boiling Capillary SS cladding fails at 1650 K Entrainment SS HP wall also fails at 1650 K 1000 Sonic TOP scenario

• The start of the fission product release Steady state occurs through the failed cladding Boiling Limit locations 100 700 800 900 1000 1100 1200 1300 1400 1500 Temperature (K) 54

Transient Overpower (TOP) base scenario (3/7)

HP Internal Gas Pressures 6

Cladding failure at 1650 K 5 resulting in fission product 4 release Pressure (bar) 3

• HPs that exceeded the boiling limit rapidly heat to cladding failure 2 (1650 K) 1

• ~20% of the 1134 HPs and fuel elements failed 0 0 1 2 3 Time (hr) 4 5 6

• HP depressurization on failure drive 1 Iodine Release and Distribution release from the vessel 0.1 Iodine releases also depend 0.01 on time at temperature Release Fraction (-)

Released In-vessel Reactor Building

• Fuel release - 1.4% of core 0.001 Environment inventory 0.0001

• Environmental release - 0.0008% of core inventory

• Vessel leakage is 1.6 in2 0.00001

• Building leakage is 1.8 in2 0.000001 0 6 12 18 24 Time (hr) 55

Transient Overpower (TOP) base scenario (4/7)

HP Internal Gas Pressures 16 The HPs could be challenged by creep failure at 14 12 high temperature and pressure 10 Pressure (bar)

• The HP gas heats and pressurizes during the TOP 8 scenario 6 Intact HPs

• The HP depressurizes after the wall fails shortly after 4 reaching the boiling limit 2 Failed HPs Creep accumulation effectively stops upon HP wall failure 0 without P stress 0 6 12 18 24 Time (hr)

HP Creep Index

• For HPs that do not reach the boiling limit, the HP 1 Creep failure at 1 pressure initially drops due to secondary system 0.1 removing heat 0.01 Creep Index (-)

HP creep failure is monitored using Larson-Miller 0.001 correlations TOP base scenario shows maximum creep is ~0.07 0.0001 (failure = 1) 0.00001 Creep failures in the condenser can create a bypass leak 0.000001 path to the environment 0 6 12 Time (hr) 18 24 56

Transient Overpower (TOP) base scenario (5/7)

Release from the fuel Fission products are retained in the fuel or deposit on their Released, 1.4%

way to the environment

• The cladding remained intact for ~80% of the fuel elements

• 98.4% of the iodine fission product inventory is retained in the Fuel fuel due to limited time at high temperature 98.6%

Distribution of Released Iodine Environment

• The vessel retains 89% of the released iodine radionuclides 0.05%

React Bldg HP depressurization after failure is primary release mechanism 11.3%

• The reactor building retains 11% of the radionuclides in the base case BWR reactor building leak tightness used for the base case No strong driving pressure to cause leakage Vessel 88.7%

57

Transient Overpower (TOP) base scenario (6/7)

A series of calculations were performed to 1.E+00 Iodine Release and Distribution investigate the sensitivity of the source term Released from fuel magnitude to reactor building leakage 1.E-01 1X Leakage 10X Leakage effects 100X Leakage

• The design specifications of the reactor 1.E-02 building were assumed Release Fraction (-)

The base result (1X) assumed a BWR reactor 1.E-03 building value 10X and 100X reflects higher design leakage 1.E-04 and/or building damage 1.E-05

• Building leakage is driven by a very small temperature gradient to the environment

(~5-7 )

1.E-06 Leakage is approximately linear with leakage 1.E-07 area (1X is ~1.8 in2) 0 6 12 18 24 Time (hr) 58

Transient Overpower (TOP) base scenario (7/7)

Iodine Release and Distribution A series of calculations were performed to 1.E+00 investigate the impact of an external wind 1.E-01

• External wind effects are included in DOE facility safety analysis where there also are not strong driving forces 1.E-02 Release Fraction (-)

Wind increases building infiltration and exfiltration 1.E-03 Upwind and downwind leakage pathways 1.E-04 Released from fuel

• Wind effects are modeled as a Bernoulli 1X Leakage, 0 mph term 1X Leakage, 5 mph 1X Leakage, 10 mph 1 1.E-05

= 2 10X Leakage, 0 mph 2 10X Leakage, 5 mph 10X Leakage, 10 mph ASHRAE building wind-pressure coefficients 1.E-06 100X Leakage, 0 mph 100X Leakage, 5 mph 100X Leakage, 10 mph 1.E-07 0 6 12 18 24 Time (hr)

External wind modeling ref:

MELCOR Computer Code Application Guidance for Leak Path Factor in Documented Safety Analysis, U.S. DOE, May 2004.

Building wind pressure coefficients.

ASHRAE, 1977, Handbook of Fundamentals, American Society of Heating, Refrigerating and Air-Conditioning Engineers, Inc, 1997. 59

Heat Pipe Reactor MELCOR Uncertainty Analysis

Role of MELCOR in Resolving Uncertainty Uncertainty Engineering Performance Event Scenario Phenomenological Uncertainty Model Uncertainty Plant Risk-Informed Assessment Initial/Boundary SSC Failure Modes Condition Uncertainty Simulation Uncertainty

Evolution from MELCOR LWR Uncertainty Analysis MELCOR application to LWR severe accident uncertainties

• Range of uncertainty studies under SOARCA

• PWR and BWR plant uncertainty studies

• Resolved role of uncertainty in a number of critical severe accident issues of high impact General commonalities between LWR and HPR accident uncertainties

• Chemical form of key elements

• Aerosol physics parameters (e.g., shape factor)

• Operating time before accident happens

• Containment leakage hole size Parameter selection emphasized potential HPR-specific uncertainties

• Ran samples of uncertainty calculations to explore role of uncertainty in evolution of HPR accident scenario class 62

Parametric Uncertainties - Capability Demonstration Component Parameter Ranges Heat Pipe Failure Location Condenser (50%) / Evaporator (50%)

Heat Pipes Initial non-functional HPs 0% - 5%

Gaseous Iodine Fraction (-) 0.0 - 0.05 Core Reactivity Insertion Rate ($/s) 0.5x10 1.0x10-3 Total reactivity feedback -0.0015 to -0.0025 Fuel Element Radial View Factor Multiplier (-) 0.5 - 2.0 Vessel Emissivity (-) 0.125 - 0.375 Vessel Total Leak Area (m2) 2x10 2x10-3 Vessel and Vessel Upper Head HTC (W/m-K) 1 - 10 Cavity entrance open fraction 100% (90%) - 1% (10%)

Cavity Emissivity (-) 0.125 - 0.375 Confinement Wind Loading (m/s) 0 - 10 Total Leak Area Multiplier (-) 1 - 100 Scenario Peak fuel temperature for safety rod insertion (K) 1300 - 2200

Characterization of Uncertainty in Event Evolution Realizations with greater Traditional event scenario evolution for LWRs reactivity insertion rates dominated by active system performance Event scenarios evolved based often on binary decisions

• SSC performance often characterized as success or failure

• Risk profile could be adequately characterized or bounded by success or failure of SSCs HPR accident scenario evolution will be unique, like other advanced non-LWRs

• Limited operational experience

• Broader range of operation for passive systems

• Consideration of degraded modes of operation

• What is the true margin to failure under accident conditions?

64

Overall Timing of Event Evolution Fission product release commences with cladding failures

• Continued fuel heatup can occur as deposited energy diffuses following reactor trip 65

Evaluating Heat Pipe Response Spectrum of accident scenarios give rise to range of plant conditions

• Relevant to assessing potential and magnitude of consequences Evaluation of SSC performance and margin in performance under accident conditions HPRs rely on passive heat removal through capillary flows in heat pipes

• Sensitive to operating range of heat pipes

• Operating limits could for example be challenged under overpower conditions 66

Fuel Response by Ring Highest powered rings off-center Energy deposited in reactor during reactivity transient diffuses to lower power rings after reactor trip Heatup of fuel in peripheral rings influenced by

• Lower decay heat levels

• Energy loss to confinement through vessel wall Heatup of fuel in central rings influenced by

• Diffusion of energy from hottest fuel rings

• Limited heat sinks to which to dissipate energy 67

Thermal Inertia in Fuel Response Most realizations dominated by early energy Diffusive heat flux from hottest rings to periphery deposition into fuel prior to reactor trip

• Dominates heatup of fuel in peripheral rings 68

Thermal Inertia in Fuel Response Centrally Peaked Core Higher powered rings off-center 69

Heat Pipe Response Lower peak fuel/clad temperatures promote potential for creep failure 70

Fission Product Release from Fuel Characterization In-vessel Iodine Release In-vessel Cesium Release Percent of Initial Inventory (%) Percent of Initial Inventory (%) 71

Fission Product Transport Characterization Reactor Building Iodine Reactor Building Cesium Percent of Initial Inventory (%) Percent of Initial Inventory (%) 72

Fission Product Release to Environment Iodine Environment Release Cesium Environment Release Percent of Initial Inventory (%) Percent of Initial Inventory (%) 73

Summary Conclusions Added HPR modeling capabilities to SCALE & MELCOR for HPR source term analysis to show code readiness Modeling demonstrated for a Transient Overpower Scenario with delayed scram

• Input of detailed ORIGEN radionuclide inventory data from ORNL

• Input radial and axial power distributions from ORNL neutronic analysis

• Develop MELCOR input model for exploratory analysis

• Fast-running calculations facilitate sensitivity evaluations (600 realizations included in the exploratory calculations)

Developed an understanding of non-LWR beyond-design-basis-accident behavior and overall plant response 75

SCALE Overview SCALE Development for Regulatory Applications What Is It?

The SCALE code system is a modeling and simulation suite for nuclear safety analysis and design. It is a modernized code with a long history of application in the regulatory process.

How Is It Used?

SCALE is used to support licensing activities in NRR (e.g., analysis of spent fuel pool criticality, generating nuclear physics and decay heat parameters for design basis accident analysis) and NMSS (e.g., review of consolidated interim storage facilities, burnup credit).

Who Uses It?

SCALE is used by the U.S. Nuclear Regulatory Commission (NRC) and in 61 countries (about 10,000 users and 33 regulatory bodies).

How Has It Been Assessed?

SCALE has been validated against criticality benchmarks (>1000), destructive assay of fuel and decay heat for PWRs and BWRs (>200) 77

Data to generate for MELCOR: QOIs 78

MELCOR for Accident Progression and Source Term Analysis

MELCOR Development for Regulatory Applications What Is It?

MELCOR is an engineering-level code that simulates the response of the reactor core, primary coolant system, containment, and surrounding buildings to a severe accident.

Who Uses It?

MELCOR is used by domestic universities and national laboratories, and international organizations in around 30 countries. It is distributed as part of NRCs Cooperative Severe Accident Research Program (CSARP).

How Is It Used?

MELCOR is used to support severe accident and source term activities at NRC, including the development of regulatory source terms for LWRs, analysis of success criteria for probabilistic risk assessment models, site risk studies, and forensic analysis of the Fukushima accident.

How Has It Been Assessed?

MELCOR has been validated against numerous international standard problems, benchmarks, separate effects (e.g., VERCORS) and integral experiments (e.g., Phebus FPT), and reactor accidents (e.g., TMI-2, Fukushima).

80

Source Term Development Process Experimental Basis PIRT process Oxidation/Gas Generation Melt Progression Fission Product Release Fission Product Transport Accident Analysis Design Synthesize MELCOR Scenario # 1 Scenario # 2 timings and Basis

. . release Source fractions Term Scenario # n-1 Scenario # n Cs Diffusivity 81

SCALE/MELCOR/MACCS Neutronics Integrated Severe Radiological SCALE MELCOR MACCS

• Criticality Accident Progression Consequences

• Shielding

• Hydrodynamics for range

• Near- and far-field

• Radionuclide inventory of working fluids atmospheric transport

• Burnup credit

• Accident response of plant and deposition

• Decay heat structures, systems and

• Assessment of health and components economic impacts

• Fission product transport Nuclear Reactor System Applications Non-Reactor Applications Safety/Risk Assessment Regulatory Design/Operational Support Fusion Spent Fuel Facility Safety

• Technology-neutral

• License amendments

• Design analysis scoping

• Neutron beam injectors

• Risk studies

• Leak path factor calculations o Experimental

• Risk-informed regulation calculations

• Li loop LOFA transient analysis

• Multi-unit accidents

• DOE safety toolbox codes o Naval

• Design certification (e.g.,

• Training simulators

• ITER cryostat modeling

• Dry storage

• DOE nuclear facilities (Pantex, o Advanced LWRs NuScale)

• He-cooled pebble test blanket

• Spent fuel transport/package Hanford, Los Alamos, o Advanced Non-LWRs

• Vulnerability studies (H3) applications Savannah River Site)

• Accident forensics (Fukushima,

• Emergency preparedness TMI)

• Emergency Planning Zone

• Probabilistic risk assessment Analysis 82

MELCOR Attributes Foundations of MELCOR Development Fully integrated, engineering-level code Phenomena modeled

• Thermal-hydraulic response of reactor coolant system, reactor cavity, rector enclosures, and auxiliary buildings

• Core heat-up, degradation and relocation

• Core-concrete interaction

• Flammable gas production, transport and combustion

• Fission product release and transport behavior Level of physics modeling consistent with

• State-of-knowledge

• Necessity to capture global plant response

• Reduced-order and correlation-based modeling often most valuable to link plant physical conditions to evolution of severe accident and fission product release/transport Traditional application

• Models constructed by user from basic components (control volumes, flow paths and heat structures)

• Demonstrated adaptability to new reactor designs - HPR, HTGR, SMR, MSR, ATR, Naval Reactors, VVER, SFP, 83

MELCOR Attributes MELCOR Pedigree International Collaboration Cooperative Severe Accident Research Program (CSARP) - June/U.S.A Validated physical models MELCOR Code Assessment Program (MCAP) - June/U.S.A European MELCOR User Group (EMUG) Meeting - Spring/Europe

• International Standard Problems, benchmarks, experiments, and reactor European MELCOR User Group (EMUG) Meeting - Fall/Asia accidents

• Beyond design basis validation will always be limited by model uncertainty that arises when extrapolated to reactor-scale Cooperative Severe Accident Research Program (CSARP) is an NRC-sponsored international, collaborative community supporting the validation of MELCOR International LWR fleet relies on safety assessments performed with the MELCOR code 84

Common Phenomenology 85

MELCOR Modeling Approach Modeling is mechanistic consistent with level of knowledge of phenomena supported by experiments Parametric models enable uncertainties to be characterized

• Majority of modeling parameters can be varied

• Properties of materials, correlation coefficients, numerical controls/tolerances, etc.

Code models are general and flexible

• Relatively easy to model novel designs

• All-purpose thermal hydraulic and aerosol transport code 86

MELCOR State-of-the-Art MELCOR Code Development Version Date 2.2.18180 M2x Official Code Releases December 2020 2.2.14959 October 2019 2.2.11932 November 2018 2.2.9541 February 2017 2.1.6342 October 2014 2.1.4803 September 2012 2.1.3649 November 2011 2.1.3096 August 2011 2.1.YT August 2008 2.0 (beta) Sept 2006

MELCOR Software Quality Assurance - Best Practices MELCOR SQA Standards Emphasis is on Automation SNL Corporate procedure IM100.3.5 Affordable solutions CMMI-4+

NRC NUREG/BR-0167 Consistent solutions MELCOR Wiki Bug tracking and reporting

• Archiving information

• Bugzilla online

• Sharing resources (policies, conventions, information, progress) Code Validation among the development team.

• Assessment calculations

• Code cross walks for complex phenomena where Code Configuration Management (CM) data does not exist.

• Subversion

• TortoiseSVN Documentation

• Available on Subversion repository with links from

• VisualSVN integrates with Visual Studio wiki (IDE)

• Latest PDF with bookmarks automatically generated from word documents under Subversion Reviews control

• Code Reviews: Code Collaborator

• Links on MELCOR wiki

• Internal SQA reviews Project Management Continuous builds & testing

• Jira for tracking progress/issues

• DEF application used to launch multiple

• Can be viewable externally by stakeholders jobs and collect results

• Regression test report Sharing of information with users

• External web page

• More thorough testing for code release

• MELCOR workshops

• Target bug fixes and new models for

• MELCOR User Groups (EMUG & AMUG) testing 88

MELCOR Verification & Validation Basis LWR & non-LWR applications Volume 1: Primer & User Guide Volume 2: Reference Manual Volume 3: MELCOR Assessment Problems

[SAND2015-6693 R]

Specific to non-Analytical Problems Saturated Liquid Depressurization Adiabatic Expansion of Hydrogen AB-1 LOF,LOHS,TOP MSRE Air-Ingress Transient Heat Flow in a Semi-Infinite Heat Slab AB-5 TREAT M-Series LWR application experiments Helical SG HT T-3 ANL-ART-38 Cooling of Heat Structures in a Fluid Radial Heat Conduction in Annular Structures Establishment of Flow Sodium Fires Molten Salt Sodium Reactors HTGR (Completed) (planned) (planned) (planned) 89

Sample Validation Cases TRISO Diffusion Release Turbulent LACE LA1 and LA3 IAEA CRP-6 Benchmark tests experimentally Deposition Fractional Release examined the Case 1a 1b 2a 2b 3a 3b transport and US/INL 0.467 1.0 0.026 0.996 1.32E-4 0.208 retention of US/GA 0.453 0.97 0.006 0.968 7.33E-3 1.00 aerosols through US/SNL 0.465 1.0 0.026 0.995 1.00E-4 0.208 pipes with high US/NRC 0.463 1.0 0.026 0.989 1.25E-4 0.207 speed flow France 0.472 1.0 0.028 0.995 6.59E-5 0.207 Korea 0.473 1.0 0.029 0.995 4.72E-4 0.210 Germany 0.456 1.0 0.026 0.991 1.15E-3 0.218 Resuspension (1a): Bare kernel (1200 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) A sensitivity study to examine STORM (Simplified Test of Resuspension (1b): Bare kernel (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) fission product release from Mechanism) test facility (2a): kernel+buffer+iPyC (1200 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) a fuel particle starting with a (2b): kernel+buffer+iPyC (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)

(3a): Intact (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) bare kernel and ending with (3b): Intact (1800 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) an irradiated TRISO particle; Aerosol Physics

• Agglomeration

• Deposition

• Condensation and Evaporation at surfaces Validation Cases

• Simple geometry: AHMED, ABCOVE (AB5 & AB6), LACE(LA4),

• Multi-compartment geometry: VANAM (M3), DEMONA(B3)

• Deposition: STORM, LACE(LA1, LA3) 90

MELCOR Modernization Generalized numerical solution engine Hydrodynamics In-vessel damage progression Ex-vessel damage progression Fission product release and transport 91