ML21179C060
| 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
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 SCALE/MELCOR Non-LWR Source Term Demonstration Project -
Heat Pipe Reactor June 2021
2 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 Outline
3 Integrated Action Plan (IAP) for Advanced Reactors Near-Term Implementation Action Plan Strategy 1 Knowledge, Skills, and Capacity Strategy 2 Analytical Tools Strategy 3 Flexible Review Process Strategy 4 Industry Codes and Standards Strategy 5 Technology Inclusive Issues Strategy 6 Communication ML17165A069
4 IAP Strategy 2 Volumes ML20030A177 ML20030A174 ML20030A176 ML20030A178 ML21085A484 Introduction Volume 1 Volume 2 Volume 3 Volume 4 Volume 5 ML21088A047 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.
5 NRC strategy for non-LWR analysis (Volume 3)
6 Role of NRC severe accident codes
Project Scope
8 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 Project objectives
9 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 Project scope
10
- 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 Project approach
11 Broad Landscape High-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 (HermeslRTR)
Terrestrial
- Thorcon Flibe TerraPower/GEH (Natrium)*
Elysium Liquid Salt Fueled TRISO Fuel Sodium-Cooled Lead-Cooled Alpha Tech Muons Micro Reactors Oklo Stationary Transportable Ultra Safe lRTR Radiant lRTR Westinghouse (eVinci)
Liquid Salt Cooled X-energy BWX Technologies Southern (TP MCFR) lRTR Oklo ARDP Awardees MIT ACU lRTR
- ARC-20 Demo Reactors In Licensing Review Risk Reduction Preapplication RTR Research/Test Reactor LEGEND General Atomics (EM2)
Kairos
- TerraPower Advanced Reactor Designs
Heat Pipe Reactor
13 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 Heat pipe for reactor use
14 Heat pipe wick being installed
15 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 First heat pipe reactor
16 LANL Megapower
- Megawatt-size heat pipe reactor
- Described in LA-UR-15-28840 Publicly available designs INL Designs A and B
- Two alternatives to Megapower for improved performance and ease of construction
- Described in INL-EXT-17-43212, Rev 1
17 LANL Megapower
- 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
18 Reactor
- 5 MW thermal power
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 INL Design A (1/2)
19 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 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 Design A (2/2)
[INL-EXT-17-43212, Rev 1]
[INL/CON-17-41817]
Heat Pipe Reactor Fission Product Inventory/Decay Heat Methods and Results
21 SCALE capabilities used KENO or Shift 3D Monte Carlo transport ENDF/B-VII.1 continuous energy physics ORIGEN for depletion Sequences
CSAS for reactivity (e.g. rod worth)
TRITON for reactor physics & depletion Relatively small amount of data except for nuclide inventory new interface file developed for inventory using standard JSON format easily read in python and post-processed into MELCOR or MACCS input contains nuclear data such as decay Q-value for traceability when performing UQ studies Workflow Power distributions Other MACCS Input MELCOR Input SCALE Binary Output Inventory Interface File SCALE Kinetics data SCALE specific Generic End-user specific SCALE Text Output
22 5 MWt rated power with 5-year operating lifetime UO2 fuel with 19.75% 235U enrichment 4.57 MTU initial core loading 1.0951 MW/MTU specific power 2.0 GWD/MTU discharge burnup 1,134 heat pipe/fuel element units Discretized with 20 axial and 5 radial fuel zones INL Design A Neutronics Summary
~100 cm Fuel Potassium heat pipe Fuel Element Lattice Cross Section at Midline Control drum 200 cm core height 3D Core Model Alumina reflector Helium void
23
- 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 HPR-Related SCALE Updates 3D SCALE model of INL Design A
24
- Example of 3D beginning-of-life (BOL) flux map overlay HPR-Related SCALE Updates Front (X-Z)
Top (X-Y) 3D Normalized flux
25
- 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])
Modeling Assumptions 3D Fission Rate
26
- 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 ENDF/B-VIII vs. ENDF/B-VII.1 comparison shows an average ~300 pcm lower
- 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 Verification and Validation 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 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
27
- 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 Unit Cell Verification Case Outer SS Clad (cm)
Pitch (cm) kinf 1 (MCNP) kinf (SCALE)
ENDF 7.0, Isotopics kinf (SCALE)
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 reflective boundary conditions1
- 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 1.
Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor, NL/EXT-17-43212, May 2018
28 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 Full Core Verification Effect of control drum rotation on eigenvalue Control Condition/Parameter Design A MCNP Design A SCALE ENDF 7.0, Isotopics Design A SCALE 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%)
29 Control Drum Rotation Flux Animations Shutdown rods in Shutdown rods out
30 Validation Basis: Short-Term Decay Heat Differential Energy Release (MeV/fission)
Fissioning nuclides in INL A 90% from 235U 10% from 238U Negligible from Pu Cumulative energy release following shutdown
~90% by 0.3 days
~92% by 1 day
~96% by 10 days 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
31 1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 0
1 2
3 4
5 6
7 8
9 10 Decay Heat (watts/MTIHM)
Time (days)
Decay heat for select isotopes for the INL Design A at 2 GWd/MTU la140 pr144 np239 i132 nb95 zr95 y91 sr89 ba140 ru103 Subtotals Totals Decay Heat after Core Shutdown
- Top 10 decay heat producing isotopes in the first 10 days following shutdown
- Subtotal shows sum of top 10
32 INL A has 2.7% specific power of PWR Comparing INL A fuel vs.
PWR fuel per MTIHM PWR at 2 GWd/MTU INL A has 2.9% of PWR decay heat at t=0 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 decay heat at t=0 INL A has 2.6% of PWR decay heat at t=10 days Does this mean decay heat can be scaled with specific power?
Why Is Decay Heat So Much Lower than a Pressurized Water Reactor (PWR)?
1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 0
1 2
3 4
5 6
7 8
9 10 Decay Heat (W/MTIHM)
Time (days)
INL A 2 GWd/MTIHM PWR 2 GWd/MTIHM PWR 60 GWd/MTIHM
33
13.2% at t=0
-10.0% at t=1.88
-5.2% at t=10
- Decay heat does not scale using specific power Scaling HPR Decay Heat Curve from a PWR
-15.0%
-10.0%
-5.0%
0.0%
5.0%
10.0%
15.0%
1.0E+03 1.0E+04 1.0E+05 0
1 2
3 4
5 6
7 8
9 10 Difference (%)
Decay Heat (watts/MTIHM)
Time (days)
INL A 2 GWd/MTIHM PWR 60 GWd/MTIHM INL A / PWR scaled -1 (%)
34 Activity after Core Shutdown 1.00E-02 1.00E-01 1.00E+00 1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 0
2 4
6 8
10 Activity (Ci/MTIHM)
Time (days)
Activity for selected Cs and I isotopes for the INL Design A at 2 GWd/MTU i131 i132 i133 cs137 i135 i134 cs138
35 Axial-normalized power peaking factors agree well with distribution from LANL and INL documents INL reference gives hottest pin power profile while the LANL and ORNL are core averages Peaks at the top and bottom are due to axial reflection Not as much reflection in the top due to heat pipes MCNP models did not use fine enough mesh to capture bottom reflector peak fully Axial peaking does not fluctuate over core lifetime due to low burnup Power Distribution 0.6 0.7 0.8 0.9 1
1.1 1.2 1.3 1.4 1.5 0
10 20 30 40 50 Normalized Peaking Factor Radial Distance from Center (cm) 0.008 years ORNL 5.000 years ORNL 0
20 40 60 80 100 120 140 160 0.5 0.6 0.7 0.8 0.9 1
1.1 1.2 1.3 Axial Height above Bottom of Active Fuel (cm)
Normalized Peaking Factor LANL Ref INL Ref 0.008 years ORNL 5.000 years ORNL
36
- 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 simultaneously Outer SS clad radial expansion Gap closure and increased pitch Alumina radial expansion Control drum drift Reactivity Feedback Effects Feedback Effect (cents/°C)
-0.1074
-0.1113 UO2 Fuel Axial Elongation
-0.0422
-0.0437 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 Axial Fuel Expansion Radial Clad Expansion
37
- 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 Neutronics Summary
MELCOR Heat Pipe Reactor Model
39 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.
MELCOR Heat Pipe Reactor Modeling: 1
40 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)
MELCOR Heat Pipe Reactor Modeling: 2 Illustrative MELCOR HP component nodalization to define MELCOR variables. Actual nodalization has more nodes.
41 MELCOR Heat Pipe Reactor Modeling: 3
- Core region modeled as a 2-D multi-ring representation Horizontal cut through core region Vertical cut through core region
- Ring-to-ring radiative exchange implemented through the generalized core heat transfer pathway modeling in MELCOR fission gas plenum region BeO reflector region lower head lower support plate outer baffle plate inner baffle plate Ring 1 - Control Rod
42 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
43 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.
HP Failure Modes Modeled in MELCOR
44 MELCOR HP failure modeling
- HP temperature excursion leads to working fluid pressurization and HP wall creep failure o Larson-Miller model used for wall failure o Subsequent response includes HP failure and depressurization
- Alternate user-specified criteria for HP 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) 0 1
2 3
4 5
6 7
8 9
10 0
200 400 600 800 1000 1200 1400 1600 Pressure (MPa)
Working Fluid Temperature (°C)
INL Design A HP Pressure versus Working Fluid Temperature
- Optional user features to dynamically control or disable HP evaporator or condenser wall heat transfer and to start the fuel cell radionuclide leakage
45 HPR model can be subdivided into an arbitrary number of rings
- Generalized input matrix for fuel 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 MELCOR cascading HP failure modeling Zone 1: 1 HP element Zone 3: 12 HP elements Zone 2: 6 HP elements Zone 4: 18 HP elements Zone modeling approach used in SFP applications for cascading fuel assembly ignition [NUREG/CR-7216]
Multiple fuel rod components in the center assembly and four peripheral assemblies
Heat Pipe Reactor Plant Model and Source Term Analysis
47 MELCOR model of INL Design A - Reactor fission gas plenum region BeO reflector region lower head lower support plate outer baffle plate inner baffle plate alumina reflector core barrel neutron shield Ring 1 - Control Rod Rings 2-15 are the active core (each ring = pitch of 1 fuel element)
Ring 1 is the control rod guide Reflector and neutron shield Evaporator (fuel elements)
Levels 3-12 Condenser (secondary heat exchanger)
Level 14 Lower reflector Levels 1-2 Level 13 Reactor modeling
- 2-D reactor nodalization 14 axial levels 15 radial rings
- 14 concentric rings of heat pipes (width of ~1 fuel assembly)
- Center ring models the emergency control rod guides
- Top and bottom reflectors are in axial levels 1 and 13
- Heat pipes transfer heat to the secondary Brayton air cycle in axial level 14
- Core region is surrounded by stainless steel shroud, alumina reflector, core barrel, and B4C neutron shield
48 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
49 FL5010 (Upper Leakage)
CV1000 (Environment)
UP UP CV5000 (Reactor Cavity)
CV5005 (Reactor Building Floor 1)
CV5010 (Reactor Building Floor 2)
FL5000 (Reactor Cavity Flow)
Natural Convection FL5005 (Reactor Cavity Flow)
Natural Convection FL5015 FL5020 FL5025 (Lower Leakage)
Ground Enclosure building nodalization LANL and INL HPR descriptions did not address the enclosure building Modeling includes internal building circulation flow paths
- Natural circulation into and out of the reactor cavity
- Natural circulation within the building Building leakage addressed parametrically
- Base leakage similar to the reactor building surrounding the BWR Mark I containment
50 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
- MELCOR can also accept proprietary performance curves when available
51 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 Description of the TOP scenario
Transient Overpower (TOP) scenario timeline Steady-State Reactivity Insertion Post-SCRAM Initialization Fuel temperature stabilizes Power increase Temperature rise Heat pipe failure Core damage Fission product release t = 0 s Tmax = 2200 K t = 24 h Radial cooling by natural processes Fission product release and transport t = -5000 s
53 1000 1200 1400 1600 1800 2000 2200 2400 0
6 12 18 24 Temperature (K)
Time (s)
Maximum Fuel Temperature 0
1 2
3 4
5 6
7 8
9 0
1 2
3 4
5 6
Power (MW)
Time (hr)
Reactor Power Transient Overpower (TOP) base scenario (1/7)
HPs hit the boiling limit The control drums start rotating at t=0 sec, which leads to an increase in the core power over 0.9 hr
- Negative fuel temperature reactivity feedback limits the rate of power increase The core steadily heats until the maximum heat flux location reaches the boiling limit
- The heat transfer rate is limited above the boiling limit, which leads to a rapid heatup rate
- The SS cladding is assumed to fail at 1650 K (just below its melting point), which starts the fission product releases into the reactor
- The reactor is assumed to trip at 2200 K Radial heat dissipation and heat loss to the reactor cavity passively cools the core
- No active heat removal (secondary system trips and isolates)
Limiting HP location hits the boiling limit Assumed manual SCRAM &
secondary isolated Passive radial heat dissipation and heat loss to the reactor cavity Control rods are inserted
54 100 1000 10000 100000 700 800 900 1000 1100 1200 1300 1400 1500 Power (W/HP)
Temperature (K)
Boiling Capillary Entrainment Sonic TOP scenario Steady state Boiling Limit Transient Overpower (TOP) base scenario (2/7)
<1 min heatup to HP & clad failure The HP performance limits at the highest heat flux location show a steady heatup to the boiling limit
- Once the boiling limit is reached, there is a rapid heatup over the next minute The fuel rapidly heats to melting conditions SS cladding fails at 1650 K SS HP wall also fails at 1650 K
- The start of the fission product release occurs through the failed cladding locations HP performance limit curves with the TOP response
55 0.000001 0.00001 0.0001 0.001 0.01 0.1 1
0 6
12 18 24 Release Fraction (-)
Time (hr)
Iodine Release and Distribution Released In-vessel Reactor Building Environment 0
1 2
3 4
5 6
0 1
2 3
4 5
6 Pressure (bar)
Time (hr)
HP Internal Gas Pressures Cladding failure at 1650 K resulting in fission product release
- HPs that exceeded the boiling limit rapidly heat to cladding failure (1650 K)
- ~20% of the 1134 HPs and fuel elements failed
- HP depressurization on failure drive release from the vessel Iodine releases also depend on time at temperature
- Fuel release - 1.4% of core inventory
- Vessel leakage is 1.6 in2
- Building leakage is 1.8 in2
56 The HPs could be challenged by creep failure at high temperature and pressure
- The HP depressurizes after the wall fails shortly after reaching the boiling limit Creep accumulation effectively stops upon HP wall failure without P stress
- For HPs that do not reach the boiling limit, the HP pressure initially drops due to secondary system removing heat HP creep failure is monitored using Larson-Miller correlations TOP base scenario shows maximum creep is ~0.07 (failure = 1)
Creep failures in the condenser can create a bypass leak path to the environment Transient Overpower (TOP) base scenario (4/7) 0.000001 0.00001 0.0001 0.001 0.01 0.1 1
0 6
12 18 24 Creep Index (-)
Time (hr)
HP Creep Index Creep failure at 1 0
2 4
6 8
10 12 14 16 0
6 12 18 24 Pressure (bar)
Time (hr)
HP Internal Gas Pressures Intact HPs Failed HPs
57 Fission products are retained in the fuel or deposit on their 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 due to limited time at high temperature
- The vessel retains 89% of the released iodine radionuclides HP depressurization after failure is primary release mechanism
- 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 Transient Overpower (TOP) base scenario (5/7)
Fuel 98.6%
- Released, 1.4%
Release from the fuel Vessel 88.7%
React Bldg 11.3%
Environment 0.05%
Distribution of Released Iodine
58 Transient Overpower (TOP) base scenario (6/7)
A series of calculations were performed to investigate the sensitivity of the source term magnitude to reactor building leakage effects
- The design specifications of the reactor building were assumed The base result (1X) assumed a BWR reactor building value 10X and 100X reflects higher design leakage and/or building damage
- Building leakage is driven by a very small temperature gradient to the environment
(~5-7 )
Leakage is approximately linear with leakage area (1X is ~1.8 in2) 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0
6 12 18 24 Release Fraction (-)
Time (hr)
Iodine Release and Distribution Released from fuel 1X Leakage 10X Leakage 100X Leakage
59 Transient Overpower (TOP) base scenario (7/7)
A series of calculations were performed to investigate the impact of an external wind
- External wind effects are included in DOE facility safety analysis where there also are not strong driving forces Wind increases building infiltration and exfiltration Upwind and downwind leakage pathways
- Wind effects are modeled as a Bernoulli term
=
1 2 2 ASHRAE building wind-pressure coefficients 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.
1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0
6 12 18 24 Release Fraction (-)
Time (hr)
Iodine Release and Distribution Released from fuel 1X Leakage, 0 mph 1X Leakage, 5 mph 1X Leakage, 10 mph 10X Leakage, 0 mph 10X Leakage, 5 mph 10X Leakage, 10 mph 100X Leakage, 0 mph 100X Leakage, 5 mph 100X Leakage, 10 mph
Heat Pipe Reactor MELCOR Uncertainty Analysis
Role of MELCOR in Resolving Uncertainty Simulation Uncertainty Plant Initial/Boundary Condition Uncertainty Event Scenario Uncertainty Phenomenological Model Uncertainty SSC Failure Modes Uncertainty Engineering Performance Risk-Informed Assessment
62 MELCOR application to LWR severe accident uncertainties
- Range of uncertainty studies under SOARCA
- 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 Evolution from MELCOR LWR Uncertainty Analysis
Parametric Uncertainties - Capability Demonstration Component Parameter Ranges Heat Pipes Heat Pipe Failure Location Condenser (50%) / Evaporator (50%)
Initial non-functional HPs 0% - 5%
Core Gaseous Iodine Fraction (-)
0.0 - 0.05 Reactivity Insertion Rate ($/s) 0.5x10 1.0x10-3 Total reactivity feedback
-0.0015 to -0.0025 Vessel Fuel Element Radial View Factor Multiplier (-)
0.5 - 2.0 Vessel Emissivity (-)
0.125 - 0.375 Total Leak Area (m2) 2x10 2x10-3 Vessel and Vessel Upper Head HTC (W/m-K) 1 - 10 Confinement Cavity entrance open fraction 100% (90%) - 1% (10%)
Cavity Emissivity (-)
0.125 - 0.375 Wind Loading (m/s) 0 - 10 Total Leak Area Multiplier (-)
1 - 100 Scenario Peak fuel temperature for safety rod insertion (K) 1300 - 2200
64 Traditional event scenario evolution for LWRs 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?
Characterization of Uncertainty in Event Evolution Realizations with greater reactivity insertion rates
65 Overall Timing of Event Evolution Fission product release commences with cladding failures Continued fuel heatup can occur as deposited energy diffuses following reactor trip
66 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 Evaluating Heat Pipe Response
67 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
68 Thermal Inertia in Fuel Response Diffusive heat flux from hottest rings to periphery Dominates heatup of fuel in peripheral rings Most realizations dominated by early energy deposition into fuel prior to reactor trip
69 Thermal Inertia in Fuel Response Centrally Peaked Core Higher powered rings off-center
70 Heat Pipe Response Lower peak fuel/clad temperatures promote potential for creep failure
71 Fission Product Release from Fuel Characterization In-vessel Iodine Release Percent of Initial Inventory (%)
In-vessel Cesium Release Percent of Initial Inventory (%)
72 Fission Product Transport Characterization Reactor Building Iodine Percent of Initial Inventory (%)
Reactor Building Cesium Percent of Initial Inventory (%)
73 Fission Product Release to Environment Iodine Environment Release Percent of Initial Inventory (%)
Cesium Environment Release Percent of Initial Inventory (%)
Summary
75 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
SCALE Overview
77 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)
78 Data to generate for MELCOR: QOIs
MELCOR for Accident Progression and Source Term Analysis
80 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).
81 Source Term Development Process Fission Product Transport MELCOR Oxidation/Gas Generation Experimental Basis Melt Progression Fission Product Release PIRT process Accident Analysis Design Basis Source Term Scenario # 1 Scenario # 2 Synthesize timings and release fractions Cs Diffusivity Scenario # n-1 Scenario # n
82 SCALE/MELCOR/MACCS Safety/Risk Assessment
- Technology-neutral o
Experimental o
Naval o
Advanced LWRs o
Advanced Non-LWRs
- Accident forensics (Fukushima, TMI)
- Probabilistic risk assessment Regulatory
- License amendments
- Risk-informed regulation
- Design certification (e.g.,
NuScale)
- Vulnerability studies
- Emergency Planning Zone Analysis Design/Operational Support
- Design analysis scoping calculations
- Training simulators Fusion
- Neutron beam injectors
- Li loop LOFA transient analysis
- ITER cryostat modeling
- He-cooled pebble test blanket (H3)
Spent Fuel
- Risk studies
- Multi-unit accidents
- Dry storage
- Spent fuel transport/package applications Facility Safety
- Leak path factor calculations
- DOE safety toolbox codes
- DOE nuclear facilities (Pantex, Hanford, Los Alamos, Savannah River Site)
Nuclear Reactor System Applications Non-Reactor Applications SCALE Neutronics
- Criticality
- Shielding
- Radionuclide inventory
- Burnup credit
- Decay heat MELCOR Integrated Severe Accident Progression
- Hydrodynamics for range of working fluids
- Accident response of plant structures, systems and components
- Fission product transport MACCS Radiological Consequences
- Near-and far-field atmospheric transport and deposition
- Assessment of health and economic impacts
83 Phenomena modeled Fully integrated, engineering-level code
- 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, MELCOR Attributes Foundations of MELCOR Development
84 Validated physical models
- International Standard Problems, benchmarks, experiments, and reactor 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 MELCOR Attributes MELCOR Pedigree International Collaboration Cooperative Severe Accident Research Program (CSARP) - June/U.S.A MELCOR Code Assessment Program (MCAP) - June/U.S.A European MELCOR User Group (EMUG) Meeting - Spring/Europe European MELCOR User Group (EMUG) Meeting - Fall/Asia
85 Common Phenomenology
86 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 MELCOR Modeling Approach
MELCOR State-of-the-Art MELCOR Code Development M2x Official Code Releases Version Date 2.2.18180 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
88 MELCOR Software Quality Assurance - Best Practices MELCOR Wiki
- Archiving information
- Sharing resources (policies, conventions, information, progress) among the development team.
Code Configuration Management (CM)
- Subversion
- TortoiseSVN
- VisualSVN integrates with Visual Studio (IDE)
Reviews
- Code Reviews: Code Collaborator
- Internal SQA reviews Continuous builds & testing
- DEF application used to launch multiple jobs and collect results
- Regression test report
- More thorough testing for code release
- Target bug fixes and new models for testing Emphasis is on Automation Affordable solutions Consistent solutions MELCOR SQA Standards SNL Corporate procedure IM100.3.5 CMMI-4+
NRC NUREG/BR-0167 Bug tracking and reporting Bugzilla online Code Validation Assessment calculations Code cross walks for complex phenomena where data does not exist.
Documentation Available on Subversion repository with links from wiki Latest PDF with bookmarks automatically generated from word documents under Subversion control Links on MELCOR wiki Project Management Jira for tracking progress/issues Can be viewable externally by stakeholders Sharing of information with users External web page MELCOR workshops MELCOR User Groups (EMUG & AMUG)
89 MELCOR Verification & Validation Basis AB-1 AB-5 T-3 Sodium Fires (Completed)
Molten Salt (planned)
Air-Ingress Helical SG HT MSRE experiments HTGR (planned)
Sodium Reactors (planned)
LOF,LOHS,TOP TREAT M-Series ANL-ART-38 Volume 1: Primer & User Guide Volume 2: Reference Manual Volume 3: MELCOR Assessment Problems Analytical Problems Saturated Liquid Depressurization Adiabatic Expansion of Hydrogen Transient Heat Flow in a Semi-Infinite Heat Slab Cooling of Heat Structures in a Fluid Radial Heat Conduction in Annular Structures Establishment of Flow Specific to non-LWR application LWR & non-LWR applications
[SAND2015-6693 R]
90 Sample Validation Cases Case 1a 1b 2a 2b 3a 3b US/INL 0.467 1.0 0.026 0.996 1.32E-4 0.208 US/GA 0.453 0.97 0.006 0.968 7.33E-3 1.00 US/SNL 0.465 1.0 0.026 0.995 1.00E-4 0.208 US/NRC 0.463 1.0 0.026 0.989 1.25E-4 0.207 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 (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 />)
(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 />)
(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 />)
(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 />)
(3b): Intact (1800 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)
IAEA CRP-6 Benchmark Fractional Release TRISO Diffusion Release A sensitivity study to examine fission product release from a fuel particle starting with a bare kernel and ending with an irradiated TRISO particle; STORM (Simplified Test of Resuspension Mechanism) test facility Resuspension LACE LA1 and LA3 tests experimentally examined the transport and retention of aerosols through pipes with high speed flow Turbulent Deposition Validation Cases
- Simple geometry: AHMED, ABCOVE (AB5 & AB6), LACE(LA4),
- Multi-compartment geometry: VANAM (M3), DEMONA(B3)
- Deposition: STORM, LACE(LA1, LA3)
Agglomeration Deposition Condensation and Evaporation at surfaces Aerosol Physics
91 MELCOR Modernization Generalized numerical solution engine Hydrodynamics In-vessel damage progression Ex-vessel damage progression Fission product release and transport