ML21256A231
| ML21256A231 | |
| Person / Time | |
|---|---|
| Issue date: | 09/14/2021 |
| From: | Jordan Hoellman NRC/NRR/DANU/UARP, Oak Ridge, Sandia, US Dept of Energy, National Nuclear Security Admin |
| To: | |
| Schaperow J | |
| References | |
| DE-NA0003525 | |
| Download: ML21256A231 (126) | |
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 -
FluorideSaltCooled High-Temperature Reactor (FHR)
September 14, 2021
2 NRC strategy for non-LWR source term analysis Project scope Overview of Fluoride-salt-cooled High-temperature Reactor (FHR)
FHR reactor fission product inventory/decay heat methods & results MELCOR molten salt models FHR plant model and source term analysis Summary Background slides
- SCALE
- MELCOR 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 - UCB Mark 1 FY22
- Molten-salt-fueled reactor - MSRE
- Sodium-cooled fast reactor - To be determined Project scope
10
- 1. Build MELCOR full-plant input model Use SCALE to provide decay heat and core radionuclide inventory
- 2. Scenario selection
- 3. Perform simulations for the selected scenario and debug Base case Sensitivity 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
FluorideSaltCooled High-Temperature Reactor (FHR)
13 Aircraft Nuclear Propulsion Program (ANP) - 1946-1961 Long-term strategic bomber operation using nuclear power ORNL developed the nuclear concept with the Aircraft Reactor Experiment (ARE)
Originally sodium cooled, but shifted to molten salt
2.5 MW molten salt-cooled reactor operated for 96-MW-hours in November 1954 Three Heat Transfer Reactor Experiments at Idaho National Laboratory to demonstrate the jet engine propulsion Aircraft Shield Test (AFT) - B-36 with an operating reactor flew 47 times over West Texas and New Mexico to study shielding (i.e., the reactor was operating but not part of the propulsion system)
Terminated due to inventing ballistic missile and supersonic aviation Molten-salt reactors (1/3)
Heat Transfer Reactor Experiment #3
The B-36 Aircraft Shield Test
![[1]](https://en.wikipedia.org/wiki/Convair_NB-36H#/media/File:NB36H-1.jpg){kind=link}
14 ORNL Molten Salt Reactor (MSR)
- AEC funded the Molten Salt Reactor Experiment (MSRE)
- Operated from 1965 to 1969
- 30 MWt
- Coolant was FLiBe molten salt
- Fuel was dissolved in coolant (molten fuel)
Molten-salt reactors (2/3)
MSRE
[ORNL-TM-0728]
MSRE Graphite Core Structure
15 UCB Mark 1 - circa 2013
- Coolant is FLiBe molten salt
- Core is TRISO fuel in a pebble-bed geometry
- Design description Technical Description of the Mark 1 Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant,
[UCBTH14002]
Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR), University of California, Berkeley, 2013.
- Used for the SCALE/MELCOR demonstration project Molten-salt reactors (3/3)
16 Reactor
- 236 MWth / 100 MWe
- Atmospheric pressure
- 600 core inlet
- 700 core outlet
- 976 kg/s core flowrate
- FLiBe molten salt coolant Core
- 470,000 fueled pebbles + 218,000 unfueled pebbles in core and defueling chute
- 180 MWd/kgHM discharge burnup
- 19.9% enrichment
- Online refueling Secondary system: gas-turbine at 18.6 bar with natural gas co-firing capability UCB Mark 1 (1/4)
UCB Mark 1 schematic
[UCBTH14002]
17 Recirculation loops
- Salt pumps in the hot well with FLiBe free surface
- 2X cross-over legs to coiled tube air heaters (CTAH)
- 2X cold legs with standpipes with free surface
- Drain tank with freeze valve UCB Mark 1 (2/4)
UCB Mark 1 schematic
[UCBTH14002]
18 Containment
- Most reactor and secondary components below-grade
- Compartmentalized building
- Low-free-volume reactor cavity with fire-brick insulation, steel liner, and concrete walls
- Shield building (above grade)
UCB Mark 1 (3/4)
Elevation view of UCB Mark 1 containment
[UCBTH14002]
19 Direct Reactor Auxiliary Cooling System (DRACS)
- 3 trains - 2.36 MW/train 236 MWt reactor
- Each train has 4 loops in series Primary coolant circulates to DRACS heat exchanger Molten-salt loop circulates to the thermosyphon-cooled heat exchangers (TCHX)
Water circulates adjacent to the secondary salt tube loop in the TCHX Natural circulation air circuit cools and condenses steam
- Start-up: Reactor coolant pump trip causes ball in valve to drop Reactor cavity cooling subsystem (RCCS) surrounds reactor cavity
- Thermal protection of the concrete UCB Mark 1 (4/4)
UCB Mark 1 DRACS
[UCBTH14002]
20 TRISO particle
- TRISO is a portmanteau of tristructural isotropic
- Kernel - 1.5 g of UCO, 200 µm radius
- Porous carbon buffer layer
- Contains 4730 TRISO particles
- 30 mm diameter
- 1 mm graphite outer shell
- TRISO particles are distributed in the carbon matrix region between the solid core and outer shell UCB Mark 1 fuel TRISO in a Fuel Pebble
![[3]](http://fhr.nuc.berkeley.edu/wp-content/uploads/2014/10/PEBBLE-SCHEMATIC-V2.png){kind=link}
Fluoridesaltcooled High-Temperature Reactor Fission Product Inventory/Decay Heat Methods and Results
22
- Objective
- Provide input for MELCOR accident simulation
Radionuclide inventory
Decay heat profile
Reactivity feedback coefficients
Reactivity from xenon transient
- Approach
- Apply SCALE to generate fuel composition for an equilibrium core
- Equilibrium core - operated for several years so the average burnups are no longer changing
- Evaluate neutronic characteristics FHR analysis with SCALE SCALE model of the UCB Mark 1 core
23 SCALE capabilities used:
Codes:
ORIGEN for depletion
KENO-VI 3D Monte Carlo neutron transport 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 Sequences:
CSAS for criticality/reactivity
TRITON for reactor physics & depletion Data: ENDF/B-VII.1 nuclear data library*
- A NUREG about Nuclear Data Assessment for Advanced Reactors summarizing the outcome of a recently concluded NRC-sponsored project is going to be published soon.
24 Neutronics overview (1/2)
FHR fuel pebble Relevant characteristics and differences to High Temperature Gas-cooled Reactors:
- Fuel:
UCO fuel in TRISO particles in fuel pebbles
TRISO particles located in shell instead of sphere
- Coolant: FLiBe salt instead of helium
- Moderator: graphite
25
- Challenges for modeling:
Tritium production in FLiBe TRISO particles with very high packing fraction in shell Fuel pebble inlet and outlet geometry Fuel and unfueled/graphite pebbles in different zones of the core
- Validation SCALE validation with HTGR experiments partially applicable*
Neutronics overview (2/2)
- F. Bostelmann, C. Celik, M. L. Williams, R. J. Ellis, G. Ilas, and W. A. Wieselquist, SCALE capabilities for high temperature gas-cooled reactor analysis, Ann. Nucl. Energy, vol. 147, p. 107673, 2020. https://doi.org/10.1016/j.anucene.2020.107673
26 UCB Mark 1 Model Description Description Value Reactor power 236 MWth UCO fuel density 10.5 g/cc Uranium enrichment 19.9 wt.%
Fuel kernel radius 0.2 mm Particle coating layer materials (starting from kernel)
Buffer/PyC/SiC/PyC Fuel particle coating layer thickness 0.100/0.035/0.035/0.035 mm Number of particles in pebble 4,730 Particle packing fraction in fuel pebble 40%
Radius of fuel pebble 1.5 cm Inner/outer radius of fuel zone 1.25/1.40 cm Number of fuel pebbles 470,000 Number of unfueled/graphite pebbles 218,000 Pebble packing fraction 60%
Core Inner reflector radius 35 cm Outer fuel pebble region radius 105 cm Outer graphite pebble region radius 125 cm Volume of active fuel region 10.4 m3 Average pebble thermal power 500 W Average pebble discharge burnup 180 GWd/MTIHM Average pebble full-power lifetime 1.40 years SCALE model developed based on:
[1] A. T. Cisneros, Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR), University of California, Berkeley, 2013.
[2] C. Andreades et al., Technical Description of the Mark 1 Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant, Berkeley, CA, UCBTH-14-002, 2014.
27 1.
Verification of multigroup physics 2.
Generation of equilibrium core 3.
Power profile and neutron spectrum 4.
Temperature feedback 5.
Decay heat 6.
1-group cross sections 7.
Tritium production 8.
Xenon reactivity Analysis areas SCALE model of the UCB Mark 1 core
28
- Comparison of multigroup (MG) calculation with continuous energy (CE) calculations for a pebble depletion problem
- Why not always run CE?
Significant modeling time: random distributions or particle arrays without permitting particle clipping Significant computation time: many cells/surfaces (consider thousands of particles) and use of CE data
- SCALEs MG approach for double-heterogeneous systems:
Two self-shielding calculations: (1) particle in graphite matrix, (2) pebble in lattice of pebbles Generation of problem-dependent cross sections for the fuel region through user-friendly input block The MG calculation is 5 times faster than the CE lattice calculation, and 24 times faster than the CE calculation with a random particle distribution
- 1. Verification of multigroup physics for UCB Mark 1 Calculation:
TRITON/KENO-VI CE and MG Depletion calculation to reach discharge burnup of 180 GWd/tHM Comparison between calculations:
k-eff, nuclide densities, runtime SCALE model a UCB Mark 1 pebble in a cube surrounded by FLiBe CE, random
~79 minutes CE, lattice 15.28 minutes MG 3.25 minutes x4.7 x24
29
- 1. Single pebble models
- 1. CE model: Random particle distribution
- 2. CE model: particle lattice (no clipping)
- 3. CE model: particle lattice (clipping)
CE-random results are average of 10 realizations All models contain the same amount of fuel
30 CZP: all materials 300K HFP: Fuel 1003K, TRISO layers 973K, graphite center 983K, outer graphite shell 957K, coolant 923K All statistical errors of the Monte Carlo calculations < 20 pcm
- 1. Single pebble initial criticality Model CZP HFP keff
[pcm]
[pcm]
CE, random no clipping 1.52539 (ref) 1.44765 (ref)
CE, lattice no clipping 1.52449
-39 1.44738
-13 CE, lattice clipping 1.51939
-259 1.44092
-323 MG 1.51986
-239 1.44426
-162 Result: MG keff calculations show good agreement with reference CE result independent of the temperature
31
- 1. Single pebble keff over the course of depletion Result: MG bias remains below 260 pcm over depletion Calculation details:
TRITON-KENO depletion of the HFP case 540.54 days at 333 MW/MTIHM Reactivity difference to CE random
32
Result: MG bias remains below 3% for relevant nuclide densities over depletion
33
- 1. MG performance summary for UCB Mark 1
- We confirmed the performance of SCALEs MG capability for double-heterogeneous systems in terms of keff and nuclide densities in a UCB Mark 1 single pebble depletion calculation
- SCALEs MG capability permits the calculation of accurate results in a much-reduced runtime (factor of 24 when compared to reference CE calculations)
34
- 2. Generation of equilibrium full core Goal: Determine fuel composition of pebbles in a full core corresponding to an equilibrium state Boundary conditions:
Pebble final discharge burnup: 180 GWd/tHM Average number of passes per pebble: 8 Average power: 333 MW/tHM Rods fully withdrawn Full core model discretization:
10 axial zones of equal volume 3 radial zones with 1/8th, 6/8th, 1/8th fractional volumes Assumptions:
All pebbles within a zone contain the same fuel composition Fuel composition within a zone represents average of individual pebbles of different passes/burnups in this zone inner middle outer
35 Fuel pebble burnup (GWd/MTIHM) in each axial zone depending on the pass through the core assuming constant axial/radial power:
- 2. Generation of isotopics for an equilibrium state 1
2 3
4 5
6 7
8 9
10 2
8 1
Pebbles of the various passes:
Re-enter pebble into core to complete 8 passes total Mix fuel compositions of these burnups to get average composition of axial zone 3 1
2 8
7 6
5 4
3
36
- 2. Approach to generate equilibrium inventory 1.
Depletion of surrogate pebbles in a core slice model to capture average spectral effects in equilibrium environment 2.
Depletion of every pebble according to its detailed power and spectral history (pass and zone in 3D core) based on average conditions from slice depletion 3.
Reconstruction of 3D core equilibrium composition according to axial/radial zones 4.
Check convergence for keff and core-average fuel composition:
stop or return to step 1 with new core-average fuel composition Outer iteration:
- 1. Constant power
- 2. 3D power map
37
- 2. Slice depletion model Why a slice and not a single pebble:
Representative moderator/fuel ratio Representative neighboring conditions (spectral effects)
Depletion model:
Slice through center of the core Depletion of surrogate pebbles surrounded by core-average fuel composition Axially reflected, radially vacuum boundary conditions Pebbles containing averaged equilibrium core fuel composition (not changing during depletion)
Depletable pebbles (always starting with fresh fuel, depleted during depletion)
Graphite pebbles
38 Outer iteration 1: convergence of keff and nuclide densities achieved after 8 inner iterations Outer iteration 2 using 3D power map showed similar convergence behavior
- 2. Keff and nuclide density convergence Outer iteration 1 using constant core power
39
- 3. Full core power profile Results:
- 1. Power peak in the lower core region in eq. core due to increasing burnup with axial height
- 2. Difference between power profiles of the two outer iterations very small with max. 6% in the lowermost zone
40 UCB Mark 1 and PBMR show a larger thermal peak compared to LWR UCB Mark 1 shows smaller fast flux due to scattering with the salt
- 3. Example fuel cell flux spectrum comparison LWR pin FHR pebble PBMR-400 pebble
41
- 3. Energy-dependent flux profile
42
- 3. 3D full core flux visualizations Fast flux, E > 0.625 eV Thermal flux, E < 0.625 eV Total flux at the axial center of the core
43
- 3. Radial flux distribution at axial core center (axial zone 5)
Fuel pebble zone Radial location normalized flux
44
- 3. Axial flux distribution in the fuel region Axial location normalized flux
45
- Isothermal temperature coefficient calculation:
keff calculations with material temperatures varying over a range of several hundred K Assuming constant temperature within material Fitting of reactivity to determine coefficient
- eff and coolant void coefficient
- 4. Reactivity coefficients Component Temperature Reactivity Coefficient at nominal temperature [pcm/K]
Salt coolant
-0.48 Fuel
-3.90 Graphite moderator
-1.10 Inner graphite reflector
+1.21 Outer graphite reflector
+0.61 Quantity Value [pcm]
541 +/- 20 Coolant void
-5094 +/- 21 Linear fit Slope from polynomial fit Nominal temperatures:
Fuel: 1003 K Salt coolant: 923 K Graphite moderator*: 973/983 K Inner graphite reflector: 873 K Outer graphite reflector: 973 K
- All carbonaceous materials in fuel pebbles
46
- 4. Isothermal temperature coefficients 2 statistical error bars are displayed a
b c
d Fuel 4.57E-02
-7.08E-05 1.59E-08 Moderator
-2.02E-03
-2.48E-05 3.88E-08 -2.16E-11 Inner graphite
-2.18E-02 2.07E-05
-7.55E-09 Outer graphite
-3.10E-02 3.49E-05
-1.31E-08
= + + 2 + 3 Linear fit:
-0.479 pcm/K
- 1. Linear fit for salt temperature coefficient
- 2. Polynomial fit or tabulated values for fuel, moderator, and graphite temperature coefficients Polynomial fit Polynomial fits
47
- 5. Generation of decay heat file for MELCOR Fuel composition files for the 8 passes for all 30 zones of the core from generation of equilibrium core Average compositions together according to zone volumes in the core to obtain core-average fuel composition 10-day decay calculation with ORIGEN, generating new composition file Generation of core-average inventory JSON file using ORIGEN composition file Conversion of JSON file to MELCOR DCH file while scaling to actual initial heavy metal mass in the core (0.705 tHM)
48
- 5. Generation of decay heat file for MELCOR Relative contribution of top fission products Relative contribution of top actinides
49
- 5. Decay heat comparisons UCB Mark 1: equilibrium core PWR: approximate end of cycle core (mixture of assemblies at burnup of 20, 40, 60 GWd/tHM)
50
- 6. Towards rapid inventory calculations with ORIGAMI UCB Mark 1 slice depletion (HFP)
Only small variation of 1-group removal cross section over depletion Small changes visible mainly in Pu-240 Purpose of 1-group cross section analysis: understand the spectral variations and their impact on 1-group cross sections which influence all inventory calculations
51
- 6. Axial variation of 1-group removal cross section Axial variation:
- Low variation within main core region
- Significant variation in inlet/outlet regions
- Opposing trends for certain nuclides, such as 239Pu vs. 240Pu Zone 1 Zone 10 1
10 Main core Main core middle
52
- 6. Radial variation of 1-group removal cross section Radial variation:
Significant radial variation for various nuclides inner middle outer 3
53 FHR uses FLiBe coolant Lithium is enriched to >99.5%
Li-7 because Li-6 is a neutron poison
- Li-6 and Li-7 react with neutrons to produce tritium 6Li + n 4He + 3H 7Li + n 4He + 3H + n Tritium is a potential radiological dose hazard
- 7. Tritium production Mass of FLiBe defined in the ORIGEN model is the total FLiBe mass in the entire system To irradiate just the FLiBe in the core at a given time, we scale the flux in our ORIGEN model based on what volume fraction of FLiBe is in the core ORIGEN flux is equal to x
TRITON
- Determine the flux spectrum and 1-group cross sections in FLiBe in this core ORIGEN
- Irradiate FLiBe using explicit flux magnitude scaled based on the fraction of system FLiBe in the core at any given time Tritium overview
54
- SCALE-predicted equilibrium value is 0.021 mol/day
- Equilibrium value from Cisneros was 0.023 mol/day
- Equilibrium is a balance between Li-6 production and destruction 9Be + n 4He + 6Li + e- +
6Li + n 4He + 3H
- The calculated behavior is consistent with established trends in the literature
- 7. Equilibrium tritium production rate 0.021 mol/day 3H t1/2 = 12.32 years
55 We ran 5,000 combinations of initial Li-7 enrichment and flux using SAMPLER to determine their impact on equilibrium tritium production Variations in initial tritium production rate are quite large and depend on flux and initial Li-7 enrichment Li-6 is a neutron poison, so FHR systems seek to enrich coolant in Li-7 Natural Li is 7.59% Li-6
- 7. Sensitivity analysis on tritium production Property Minimum Value Maximum Value Flux (n/cm2-s) 3.528x1014 4.312x1014 Initial Li-7 Enrichment (w/o) 99.95 100.0
56 Initial Li-7 enrichment has no effect on equilibrium tritium production rate, while flux has a significant impact
- 7. Sensitivity analysis on tritium production No correlation for initial Li-7 enrichment Strong correlation for neutron flux
57
- Steady-state Xe-135 reactivity worth is
-6.48$
- Using equilibrium I-135 and Xe-135 concentrations from UCB Mark 1 model, we can calculate time-dependent concentrations analytically
- When flux goes to zero, Xe-135 inventory is dictated only by decay of I-135 and Xe-135
- Peak Xe-135 reactivity is -18.6$ and occurs at 9.49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br />
- Xe-135 reactivity drops below steady-state value after 34.67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br />
=
= +
58 Demonstrated SCALEs capabilities for FHR modeling SCALEs multigroup physics was confirmed adequate through FHR fuel pebble analysis: keff bias smaller than 260 pcm, while achieving 24 times faster runtime Fuel compositions for an equilibrium core were developed using an iterating scheme Power profiles and decay heat were determined for equilibrium core Temperature feedback: linear behavior found for salt, nonlinear trend for fuel and for materials containing graphite Strong radial variation for 1-group cross section was observed, while axial variation was limited to inlet/outlet regions Tritium production rate in coolant salt was estimated Preliminary results for time-dependent Xe-135 concentration Neutronics Summary
MELCOR Molten Salt Models
60 Added molten salt as working fluid Fission product release
- Release from TRISO kernel
- Radionuclide distributions within the layers in the TRISO particle and compact
- Liquid-phase fission product chemistry and transport model Additional core models
- Graphite oxidation
- Intercell and intracell conduction
- Convection & flow Fluid point kinetics (liquid-fueled molten salt reactors)
MELCOR Molten Salt Reactor Modeling
61 Stage 1:
Normal Operation Diffusion Calculation Establish steady state distribution of radionuclides in TRISO particles, and matrix Stage 2:
Normal Operation Transport Calculation Calculate steady state distribution of radionuclides into the molten salt (formation of soluble, colloidal fission products, deposition on surfaces, convection through flow paths)
Stage 3:
Accident Diffusion & Transport calculation Calculate accident progression and radionuclide release Stage 0:
Normal Operation Establish thermal state Time constant in FHR graphite structures is very large Reduce heat capacities for structures to reach steady state thermal conditions.
Reset heat capacities after steady state is achieved.
0 50 100 150 200 250 300 350
-15000
-12500
-10000
-7500
-5000
-2500 0
Power (MW)
Time (sec)
Core Power CTAH HX UCB Reference 1.E-15 1.E-14 1.E-13 1.E-12 1.E-11
-1200
-1000
-800
-600
-400
-200 Fraction of initial inventory (-)
Time (sec)
Cesium Release from the Pebbles 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 0
12 24 36 48 Fraction of initial inventory (-)
Time (h)
Cesium Release from the Pebbles Transient/Accident Solution Methodology 0
50 100 150 200 250 300
-2.0
-1.5
-1.0
-0.5 0.0 0.5 1.0 1
10 100 1000 Power (MW)
Reactivity ($)
Time (sec)
Core Reactivities Fuel Temperature Molten Salt Inner Reflector Outer Reflector Moderator Xenon Total Reactivity Power
62
- Pebble Bed Reactor Fuel/Matrix Components Fueled part of pebble Unfueled shell (matrix) is modeled as separate component Fuel radial temperature profile for sphere
- Prismatic Modular Reactor Fuel/Matrix Components Rod-like geometry Part of hex block associated with a fuel channel is matrix component Fuel radial temperature profile for cylinder Core components Legend TRISO (FU)
Fuel (FU)
Matrix (MX)
Fluid B/C TRISO GRAPHITE Sub-component model for zonal diffusion of radionuclides through TRISO particle GRAPHITE Fuel Compact Unfueled Fueled pebble core Unfueled pebble core
63 Intact TRISO Particles
- One-dimensional finite volume diffusion equation solver for multiple zones (materials)
- Temperature-dependent diffusion coefficients (Arrhenius form)
Radionuclide Diffusion Release Model Intact TRISO Concentrations
= 1
+
Layer FP Species Kr Cs Sr Ag D (m2/s)
Q (J/mole)
D (m2/s)
Q (J/mole)
D (m2/s)
Q (J/mole)
D (m2/s)
Q (J/mole)
Kernel (normal) 1.3E-12 126000.0 5.6-8 209000.0 2.2E-3 488000.0 6.75E-9 165000.0 Buffer 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 PyC 2.9E-8 291000.0 6.3E-8 222000.0 2.3E-6 197000.0 5.3E-9 154000.0 SiC 3.7E+1 657000.0 7.2E-14 125000.0 1.25E-9 205000.0 3.6E-9 215000.0 Matrix Carbon 6.0E-6 0.0 3.6E-4 189000.0 1.0E-2 303000.0 1.6E00 258000.0 Str. Carbon 6.0E-6 0.0 1.7E-6 149000.0 1.7E-2 268000.0 1.6E00 258000.0 Data used in the demo calculation
[IAEA TECDOC-0978]
= 0
Diffusivity Data Availability Radionuclide UO2 UCO PyC Porous Carbon SiC Matrix Graphite TRISO Overall Ag Some Not investigated Some Not found Extensive Some Extensive Cs Some Some Extensive Some Some I
Some Some Some Not found Not found Kr Some Some Not found Some Some Sr Some Some Extensive Some Some Xe Some Some Some Some Not found Iodine assumed to behave like Kr
64
- Recent failures - particles failing within latest time-step (burst release, diffusion release in time-step)
- Previous failures - particles failing on a previous time-step (time history of diffusion release)
- Contamination and recoil Radionuclide Release Models Failing Intact TRISO Released to the matrix Transition from Intact-to-failed Failed TRISO Contamination Release from failed TRISO (Modified Booth)
Intact TRISO Failed TRISO recoil Released to the matrix Transfer to failed TRISO Distribution calculated from diffusion model Release from TRISO failure Diffusion Diffusion from intact TRISO Recoil fission source recoil Diffusion Diffusion
65 Salt vapor or He Steam oxidation Graphite Oxidation Reactions Air oxidation Reactions Air diffusion towards oxidation surface is rate limited due to mass transfer limitations in presence of salt vapor Air ROX is the rate term in the parabolic oxidation equation [1/s]
Existing capability introduced with High-Temperature Gas-cooled Reactors (HTGRs)
66
- Zehner-Schlunder-Bauer, without radiation heat transfer Effective conductivity prescription for pebble bed (bed conductance)
Energy Transport between Discrete Core Volumes
= 1 1 +
1,,,
where:
= Bed porosity [-]
= Fluid (FLiBe) conductivity [W/m/K]
= Effective bed conductivity [W/m/K], used with zero radiative conductivity
= Solid conductivity [W/m/K]
= Solid temperature [K]
0 3
6 9
12 15 0.4 0.45 0.5 0.55 0.6 Effective Conductivity [W/m/K]
Packed Bed Porosity [-]
ZSB w/o Radiation Terms ZSB w/ Radiation ks = 35.5 W/m/K kf = 1.1 W/m/K kr (T=978 K, Dp=0.03 m) = 6.3 W/m/K
=0.8 Effective fluid conductivity combines liquid and vapor contributions according to vapor fraction Radiative conductivity is combined by vapor fraction and used in ZSB model with radiation terms
= 1 1 + 1 1 +
1,,,,,
= 43
67 Heat transfer coefficient (Nusselt number) correlations for pebble bed convection:
- Isolated, spherical particles
- Use Tfilm to evaluate non-dimensional numbers, use maximum of forced and free Nu
- Constants and exponents accessible by sensitivity coefficient Interface Between Thermal Hydraulics and Reactor Core Structures Flow resistance
- Packed bed pressure drop
, = 1 + 2 1
+ 3 1
4 1
Loss coefficient relative to Ergun (original) coefficient at Re=1000
= 2.0 + 0.6
1 4
1 3
= 2.0 + 0.6
1 2
1 3
68 Standard treatment Feedback models
- User-specified external input
- FHR example includes multiple feedbacks
- Fuel
- Molten salt around the fuel
- Inner reflector
- Outer reflector and unfueled pebbles
- Moderator (matrix around fueled pebbles)
Point Kinetics Modeling
=
+
=1 6
+ 0
=
= 1 6 0
50 100 150 200 250 300
-2.0
-1.5
-1.0
-0.5 0.0 0.5 1.0 1
10 100 1000 Power (MW)
Reactivity ($)
Time (sec)
Core Reactivities Fuel Temperature Molten Salt Inner Reflector Outer Reflector Moderator Xenon Total Reactivity Power
69 Derived from standard PRKEs and solved similarly Feedback models
- User-specified external input
- Doppler
- Fuel and moderator density
- Flow reactivity feedback effects integrated into the equation set Point Kinetics Modeling (MSR)
=
+
=1 6
+ 0
=
+
2
+
+
2
= 1 6
=
+
1
= 1 6
=1 6
0 50 100 150 200 250 0
10 20 30 40 50 60 70 Compensating Control System Reactivity [pcm]
Time [s]
Guo Code MSRE Data MELCOR Validated against MSRE zero-power flow experiments
70 Molten Salt Chemistry and Radionuclide Release Radionuclides grouped into forms found in the Molten Salt Reactor Experiment MELCOR-provided state Atmospheric Release Mechanisms Evaluation of thermochemical state
- Gibbs Energy Minimization with Thermochimica
- Provides solubilities and vapor pressures Thermodynamic database
- Generalized approach to utilize any thermodynamic database
- An example is the Molten Salt Thermal Database FLiBe-based systems Chloride-based systems Solubility determined from empirical evidence (P. Britt ORNL 2017)
Solubilities mapped to 17 MELCOR fission product classes Insoluble MELCOR classes are assigned to be colloidal Model Scope Initial Model Form
Fluoridesaltcooled High-Temperature Reactor Plant Model and Source Term Analysis
72 Core and reactor vessel Core nodalization - light blue lines
- Assumes azimuthal symmetry
- Subdivided into 11 axial levels and 8 radial rings
- Core cells model molten salt fluid volume, reflector structures, the pebble-bed core, and the pebbles in the defueling chute Fluid flow nodalization - black boxes
- Molten salt enters through the downcomer and flows into the center reflector and into the bottom of the pebble bed
- Molten salt leaves through the periphery of the core and upwards through the refueling chute
- Unfueled graphite pebbles in box labeled 180
73 Recirculation loops Each loop has a pump, a heat exchanger, and a standpipe Molten salt has free surface in the hotwell and the standpipes Argon gas above the free surfaces with connection to the cover-gas system
- Over-pressurization relief passes through the cover gas system
- Cover gas enclosure leaks into the containment when over-pressurized Secondary-side air cools primary-side molten salt
74 Direct Reactor Auxiliary Cooling System (DRACS) 3 trains - 2.36 MW/train
- 236 MWt reactor Each train has 4 loops in series
- Primary coolant circulates to DRACS heat exchanger
- Molten-salt loop circulates to the thermosyphon-cooled heat exchangers (TCHX)
- Water circulates adjacent to the secondary salt tube loop in the TCHX
- Natural circulation air circuit cools and condenses steam Start-up: RCS-pump trip causes ball in valve to drop Additional system information
- DHXs are in the reactor vessel
- TCHXs are in the shield building
75 Containment Shield dome
- Protection against aircraft and natural gas detonations (co-fired turbine concept)
- Contains water for DRACS and RCCS
- DRACS air natural circulation chimneys connected to the shield dome Reactor cavity
- Fire-brick insulation
- Low free volume
- Low-leakage bellows between reactor cavity and adjacent cavities Separate compartments for the other RCS components
- Below-grade compartment includes the cover-gas enclosure for reactor cavity over-pressurization Reactor cavity cooling subsystem in reactor cavity wall
- Water circulation
- Cooling tubes affixed to reactor cavity steel liner
- Cools concrete during normal operation Leak rate assumed consistent with BWR Mark 1 reactor building
- 100% vol/day at 0.25 psig
76 MELCOR model inputs (1/2)
Equilibrium inventory and decay heat from SCALE Radial and axial power profiles from SCALE Reactivity feedbacks from SCALE Cell-to-cell radial and axial heat transfer in the pebble bed and to adjacent reflector structures
- Modified Zehner-Schlunder-Bauer model formulation
- Combined conductive and radiative (when core uncovered) heat transfer depends on the coolant and fuel conductivities, fuel (graphite) emissivity, pebble bed porosity Pebble bed friction losses - Achenbach pressure drop formulation
- = 2 + 320 (1 )
+ 20 (1 )
0.4 Pebble to fluid heat transfer within a cell
- Forced convection using Wakao correlation, Nu = 2 + 1.1 Re 0.66Pr 0.33
77 MELCOR model inputs (2/2)
Fission product diffusivities through the TRISO and the pebble matrix from IAEATECDOC978, Appendix A
- Primarily based on values from German experiments with UO2 TRISO pebbles UO2 data can be easily updated to UCO data*
- Limited data based on nuclides of Xe, Cs, Sr, and Ag
- Iodine assumed to behave like Kr 1.E-20 1.E-19 1.E-18 1.E-17 1.E-16 1.E-15 1.E-14 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 3
4 5
6 7
8 Diffusivity (m2/s) 10,000/T (K-1)
Kr Diffusivities vs TRISO Layer IAEA UO2 IAEA Buffer IAEA Pyrolytic IAEA SiC IAEA Matrix 1.E-20 1.E-19 1.E-18 1.E-17 1.E-16 1.E-15 1.E-14 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 3
4 5
6 7
8 9
Diffusivity (m2/s) 10,000/T (1/K)
Ag Diffusivities vs TRISO Layer IAEA UO2 IAEA Buffer IAEA Pyrolytic IAEA SiC IAEA Matrix
- UCO TRISO thermal failure characteristics were not available, so UO2 TRISO diffusivity and UO2 failure data were used. Both are changeable through user input with design-specific data.
78 Three scenarios with a loss of secondary heat removal
- SBO - Station blackout
- LOCA - Loss-of-coolant accident Sensitivity calculations included
- DRAC performance
- Alternate cover-gas system interconnections (LOCA only)
Scenarios
79 Loss-of-onsite power with failure to SCRAM
- Salt pumps shut off
- Reactor fails to SCRAM
- Secondary heat removal ends
- Guided by ORNL calculations
80 1
10 100 1000 1
10 100 1000 10000 100000 Power (MW)
Time (sec)
Core Energy Balance Core Power Decay Heat DRACS
-12
-10
-8
-6
-4
-2 0
2 1
10 100 1000 10000 100000 Reactivity ($)
Time (sec)
Core Reactivities Fuel Temperature Molten Salt Inner Reflector Outer Reflector Moderator Xenon Total Reactivity ATWS with 3xDRACS Initial fuel heatup has strong negative fuel and moderator feedback that offsets positive reflector feedbacks Strong negative xenon transient feedback
- 3xDRACS exceeds core power after 330 s
-1 0
1 2
85000 95000 105000 115000 125000 135000 Reactivity ($)
Time (sec)
Core Reactivities Fuel Temperature Molten Salt Inner Reflector Outer Reflector Moderator Xenon Total Reactivity 0
5 10 15 20 25 85000 95000 105000 115000 125000 135000 Power (MW)
Time (sec)
Core Energy Balance Core Power Decay Heat DRACS
81 0
5 10 15 20 25 1
10 100 1000 10000 100000 Power (MW)
Time (sec)
Core Power abd DRACS Heat Removal 3xDRACS 2xDRACS 1xDRACS 0xDRACS 3xDRACS Core Power 2xDRACS Core power 1xDRACS Core Power 0xDRACS Core Power ATWS with variable DRACS (semi-log)
Early power decrease to decay heat level is similar for all cases
- 1xDRACS and 2xDRACS cases exceed decay heat later Fission power starts increasing Fuel temperatures cool down according to DRACS heat removal rate
- 0xDRACS peak fuel temperature = 990 at 105 s (Tsat~ 1350 )
Core power DRACS heat removal 500 550 600 650 700 750 800 850 900 950 1000 1050 1
10 100 1000 10000 100000 Temperature (deg-C)
Time (sec)
Peak Fuel Temperatures 0xDRACS 1xDRACS 2xDRACS 3xDRACS Core power and DRACS Heat Removal
82 ATWS with variable DRACS - (Linear scale)
When the total reactivity exceeds zero, the core power increases
- Increased power heats fuel and reduces the positive fuel reactivity
- Core power eventually converges on the DRACS heat removal rate Fission power starts increasing Long-term fuel temperatures are similar
- Fuel temperatures and reactivity adjusts to offset changes in xenon poisoning
0 5
10 15 20 25 85000 95000 105000 115000 125000 135000 Power (MW)
Time (sec)
Core Power and DRACS Heat Removal 3xDRACS 2xDRACS 1xDRACS 3xDRACS Core Power 2xDRACS Core power 1xDRACS Core Power 500 550 600 650 700 750 800 85000 95000 105000 115000 125000 135000 Temperature (deg-C)
Time (sec)
Peak Fuel Temperatures 1xDRACS 2xDRACS 3xDRACS
83 Loss-of-onsite power with SCRAM
- Salt pumps shut off
- Reactor scrams
- Secondary heat removal ends
- Variable DRACS operating (percentage of 1xDRACS)
Unmitigated sensitivity case
- No DRACS and extended calculation to 7 days Station Blackout
84 SBO results (1/3)
DRACS cases illustrate degraded response
- Results for fraction of 1xDRACS
- >40% of one DRACS stops the temperature rise within 48 hr 0
250 500 750 1000 1250 1500 0
6 12 18 24 30 36 42 48 Temperature (oC)
Time (hr)
Peak Fuel Temperature Tsat at Core Outlet 1.0 x DRACS 0.8 x DRACS 0.6 x DRACS 0.4 x DRACS 0.2 x DRACS No DRACS 0
2 4
6 8
10 0
12 24 36 48 Power (MW)
Time (hr)
Decay Heat and DRACS Heat Rejection Decay Heat Power 1.0 x DRACS 0.8 x DRACS 0.6 x DRACS 0.4 x DRACS 0.2 x DRACS No DRACS DRACS power follows heat removal requirements
- 1xDRACS exceeds decay heat within 3 hr
85 SBO results (2/3)
The TRISO failure fraction remains low (1x10-5) in the LOPA with one DRACS operating *
- Higher TRISO failures were calculated as the DRACS degrades 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 30 36 42 48 Fraction of initial inventory (-)
Time (hr)
TRISO Failure Fraction 1.0 x DRACS 0.8 x DRACS 0.6 x DRACS 0.4 x DRACS 0.2 x DRACS No DRACS
- UCO TRISO thermal failure characteristics were not available, so UO2 TRISO diffusivity and UO2 failure data were used. Both are changeable through user input with design-specific data.
86 0
2 4
6 8
10 12 0
24 48 72 96 120 144 168 Level (m)
Time (hr)
Downcomer Level Downcomer Hotwell Top of the refueling chute SBO results (3/3)
The SBO with no DRACS was extended to 7 days
- No fuel uncovery
- Peak fuel temperature approximately at Tsat (~1350 )
Liquid salt exiting the cover gas system Liquid level is approaching the top of the refueling chute Rupture disk opens 120 hr to boiling conditions 0
200 400 600 800 1000 1200 1400 1600 0
24 48 72 96 120 144 168 Temperature ()
Time (hr)
Peak Fuel Temperature Tsat at core exit Peak fuel temperature
87 Loss-of-onsite power with LOCA
- Variable size leaks of the 3 pipe of the drain tank line
- Salt pumps shut off
- Reactor scrams
- Secondary heat removal ends
- 1 or no trains of DRACS operating
- With or without a cover gas connection path between the hotwell and the standpipes Unmitigated sensitivity case
- No DRACS case extended to include fuel uncovery LOCA
88 0
1 2
3 4
5 6
7 8
9 10 0
12 24 36 48 Level (m)
Time (hr)
Downcomer Level 100% LOCA + CG 100% LOCA 75% LOCA 50% LOCA 25% LOCA 10% LOCA LOCA results (1/6) 10% to 100% LOCA size did not significantly impact vessel boiloff timing Cover gas connection (+ CG) between hotwell and standpipe prevents siphon
- Stops initial drain down of vessel fluid
- No significant impact on vessel boiloff timing Siphon effect drains vessel until broken 0
2 4
6 8
10 12 0
3 6
9 12 Level (m)
Time (hr)
Standpipe Level 100% LOCA 75% LOCA 50% LOCA 25% LOCA 10% LOCA No drain down with cover gas connection
89 LOCA results (2/6)
Liquid drain down initially creates siphon and then low pressure region
- Causes a level difference between the core and downcomer Core and downcomer levels equilibrate once there is gas flow around the loop
- Standpipe connections to the cover gas system are closed 10% LOCA at maximum point in the siphon 10% LOCA after equilibration Gas flow
90 LOCA results (3/6)
LOCA cases without DRACS proceed to fuel uncovery at ~31 hr Connection through the cover gas system keeps the DRACS active during the drain down
- Without the cover gas connection, the DRACS heat removal is delayed until the salt heats and expands 0
1 2
3 4
5 6
7 8
9 10 0
12 24 36 48 Level (m)
Time (hr)
Downcomer Level Top of the refueling chute 100% LOCA + CG + 1xDRACS 100% LOCA + 1xDRACS 100% LOCA + CG 100% LOCA No initial drain down with cover gas connection DRACS prevents boiloff 0
250 500 750 1000 1250 1500 1750 2000 0
24 48 72 96 120 144 168 Temperature (oC)
Time (hr)
Peak Fuel Temperature Tsat at Core Outlet 100% LOCA + 1xDRACS + CG 100% LOCA + 1xDRACS 100% LOCA + CG 100% LOCA DRACS provides heat removal 100% LOCA cases 100% LOCA cases
91 LOCA results (4/6)
We terminated the calculation at ~54 hr peak when the fuel kernel melting starts
- Reactor vessel wall and core barrel below steel melting temperature
- Residual molten salt keeps the bottom level (level 1) at Tsat
- Upper vessel wall cools after downcomer salt level drops
- Pebbles and reflectors below graphite sublimation temperature (3600)
Conditions at 54 hr 0
500 1000 1500 2000 2500 3000 0
6 12 18 24 30 36 42 48 54 Temperature (oC)
Time (hr)
Peak fuel temperature Steel melting temperature Core Barrel Level 1 Core Barrel Level 6 Core Barrel Level 11 Reactor Vessel Level 1 Reactor Vessel Level 6 Reactor Vessel Level 11
92 LOCA results (5/6)
Note:
- Fuel used thermal-physical properties of UO2.
Low failure rate when <Tsat TRISO failure rate extrapolated from available UO2 TRISO data
- Correlation is based on data to 1800
- Initial failures set to 10-5 (0.001%)
- 0.017% of the TRISOs failed at 34 hr
- 7.5% of the TRISOs failed at 54 hr 500 1000 1500 2000 2500 3000 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0
6 12 18 24 30 36 42 48 54 Temperature ()
Fraction of initial inventory (-)
Time (hr)
TRISO Failure Fraction and Peak Fuel Temperature Failure fraction Peak fuel temperature Conditions at 34 hr
93 LOCA results (6/6)
Most of the fission product release from fuel is retained in the containment
- Assumed hole size equivalent to 100%
volume per day at 0.25 psig (8.7 in2)
The radionuclide distribution is affected by the timing of the release from the TRISO
- Cesium release from the pebbles to the liquid molten salt starts earlier at lower fuel temperatures
- Most aerosols leaving the primary system settle in the containment 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 30 36 42 48 54 Fraction of initial inventory (-)
Time (hr)
Iodine Release and Distribution Released Primary system Containment Envronment 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 30 36 42 48 54 Fraction of initial inventory (-)
Time (hr)
Cesium Release and Distribution Released Primary system Containment Envronment
94 Cesium vaporization from the molten salt Molten salt chemistry and radionuclide release model calculates cesium and cesium fluoride release to the gas spaces
- Results use OECD/NEA JRC database for Thermochimica *
- Includes vapor phase data for CsF LOCA sequence
- No accelerated steady state
- No core uncovery through 24 hr Cesium releases are from pebbles liquid gas Model shows Cs/CsF vaporization to gas spaces at higher temperatures
- With modifications by Ontario Tech.
600 700 800 900 1000 1100 1200 1300 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 0
6 12 18 24 Temperature (deg-C)
Fraction of initial invenory (-)
Time (hr)
Cesium Behavior Total released from pebbles Vaporized from the liquid Total in the liquid Core Fluid Temperature
Summary
96 Conclusions Demonstrated use of SCALE and MELCOR for FHR safety analysis Simulated the entire accident starting with the initiating event system thermal hydraulic response fuel heat-up heat transfer through the reactor to the surroundings radiological release Evaluated effectiveness of passive mitigation features
=
Background===
Slides
Further SCALE analysis details
99 Comparison of the FHR with other concepts C. Andreades et al., Technical Description of the Mark 1 Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant, Berkeley, CA, UCBTH-14-002, 2014.
100
- 1. Single pebble nuclide density over depletion CE random results:
101
Result: MG bias remains below 3% for relevant nuclide densities over depletion
102
- 1. Single pebble nuclide density comparison of against reference CE random results CE, lattice, unclipped CE, lattice, clipped
103 Monte Carlo calculation settings:
- 25,000 neutrons per cycle in 500 active and 100 inactive generations
- 1 node with 32 processors
- 1. Single pebble runtime comparison Model Runtime [min]
CE, random no clipping
~79 (per realization)
CE, lattice no clipping 15.28 CE, lattice clipping 15.78 MG 3.25 SCALE model a UCB Mark 1 pebble in a cube surrounded by FLiBe
104
- 2. Generation of isotopics for an equilibrium state
- Outer iteration 1:
Flat axial power profile Consider only axial zones No radial zones or radial power distribution
- Outer iteration 2:
Use axial and radial power profile from outer iteration 1 Consider axial and radial zones Additional assumption:
homogenization of compositions of all radial zones after each pass initial composition for next pass
105
- 2. Convergence of results during iterations Convergence after 8 or 9 iterations:
keff converged Nominal discharge burnup achieved Nuclide densities converged
106
- 2. Comparison of final core average fuel compositions Nuclide Density [at/b-cm]
Relative difference Outer iteration 1 Outer iteration 2 xe-135 4.587E-08 4.422E-08
-3.6%
cs-134 1.542E-05 1.509E-05
-2.2%
cs-137 1.570E-04 1.568E-04
-0.1%
nd-148 4.405E-05 4.414E-05 0.2%
sm-149 4.019E-07 4.122E-07 2.6%
sm-151 1.856E-06 1.861E-06 0.3%
gd-154 6.098E-08 6.104E-08 0.1%
gd-155 2.865E-09 3.137E-09 9.5%
eu-153 1.077E-05 1.065E-05
-1.1%
eu-154 1.788E-06 1.759E-06
-1.6%
eu-155 5.965E-07 5.876E-07
-1.5%
Nuclide Density [at/b-cm]
Relative difference Outer iteration 1 Outer iteration 2 u-235 2.316E-03 2.306E-03
-0.4%
u-238 1.786E-02 1.788E-02 0.1%
pu-239 2.127E-04 2.143E-04 0.7%
pu-240 8.041E-05 8.033E-05
-0.1%
pu-241 6.724E-05 6.662E-05
-0.9%
pu-242 2.980E-05 2.910E-05
-2.3%
am-241 6.746E-07 6.873E-07 1.9%
cm-242 4.772E-07 4.672E-07
-2.1%
cm-244 1.467E-06 1.420E-06
-3.2%
Relative difference of core-average fuel composition is negligible besides very few exceptions in case of small nuclide densities.
107
- 3. Full core power profile Results:
Power is peaking in the inner fuel region Consideration of axial/radial power profile in the iterations to obtain the equilibrium core compositions has minor effect.
108 Reactivity coefficient calculation:
keff calculations with material temperatures varying over a range of several hundred K Assuming constant temperature within material Fitting of to determine coefficient
- 4. Comparison of isothermal temperature coefficients Component Temperature Reactivity Coefficient at HFP [pcm/K]
Cisneros [1]
ORNL Fuel
-3.8
-3.90 Salt coolant
-1.8
-0.48 Graphite moderator
-0.7
-1.10 Inner graphite reflector
+0.9
+1.21 Outer graphite reflector
+0.9
+0.61
[1] A. T. Cisneros, Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR), University of California, Berkeley, 2013.
109
- 6. Towards rapid inventory calculations with ORIGAMI PBMR-400 slice depletion*
UCB Mark 1 slice depletion (HFP)
- S. Skutnik, W. Wieselquist, ORNL/TM-2020/1886, 2021. https://www.osti.gov/servlets/purl/1807271 Only small variation of 1-group removal cross section over depletion Small changes visible mainly in Pu-240 Purpose of 1-group cross section analysis: understand the spectral variations and their impact on 1-group cross sections which influence all inventory calculations
110
- 6. Comparison between UCB Mark 1 and PBMR-400 PBMR-400*
UCB Mark 1 Both cores showed significant radial variation for various nuclides Only UCB Mark 1 showed axial variation due to inlet/outlet geometry
- S. Skutnik, W. Wieselquist, ORNL/TM-2020/1886, 2021. https://www.osti.gov/servlets/purl/1807271 Axial variation provided as error bars K
111 A simplified analytical model was developed by Cisneros et al*. using a flux and one-group cross sections to allow estimation of tritium generation rates for an arbitrary initial Li-7 enrichment
- 7. Analytical model to calculate tritium production
= 7
7 + 6
6 0
6
+ 9
9 6
1
6
- Cisneros, A. T., 2013. Pebble Bed Reactors Design and Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR) (PhD). University of California Berkeley.
SCALE results using TRITON/ORIGEN: 0.021 mol/day Equilibrium value from Cisneros analytical approach: 0.023 mol/day
MELCOR for Accident Progression and Source Term Analysis
113 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).
114 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
115 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
116 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
117 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
118 Common Phenomenology
119 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
121 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)
122 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]
123 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
124 MELCOR Modernization Generalized numerical solution engine Hydrodynamics In-vessel damage progression Ex-vessel damage progression Fission product release and transport
125 Molten Salt Chemistry and Radionuclide Release -
Integration into MELCOR MELCOR provides mass of radionuclides released into salt, chemistry, T and P Chemistry model computes what remains in salt as soluble, colloidal, deposited, and released as vapor and aerosol MELCOR continues to transport materials to and from the salt control volume Each Timestep
Cs vapor pressures in MSM calculations