ML22255A214

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Presentation - 9/13/22 Scale Melcor Molten Salt Reactor (MSR) Workshop
ML22255A214
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Issue date: 09/12/2022
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Office of Nuclear Reactor Regulation, Oak Ridge
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Schaperow J
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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 - Molten Salt Reactor (MSR)

September 13, 2022 SAND2022-12146 PE

2 NRC strategy for non-LWR source term analysis Project scope Overview of Molten Salt Reactor (MSR)

MSR reactor fission product inventory/decay heat methods & results MELCOR molten salt models MSR plant model and source term analysis Summary 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 and sample calculations for representative non-LWRs 2021 Heat pipe reactor - INL Design A Pebble-bed gas-cooled reactor - PBMR-400 Pebble-bed molten-salt-cooled - UCB Mark 1 Public workshop videos, slides, reports at advanced reactor source term webpage 2022 Molten-salt-fueled reactor - MSRE - public workshop 9/13/2022 Sodium-cooled fast reactor - ABTR - public workshop 9/20/2022 2023 Additional code enhancements and sample calculations Project scope

10

1. Build SCALE core model and MELCOR full-plant model
2. Select scenarios that demonstrate code capabilities
3. Perform simulations Use SCALE to model decay heat, core radionuclide inventory, and reactivity feedback Use MELCOR to model accident progression and source term Perform sensitivity cases Project approach

Molten Salt Reactor (MSR)

12 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 reactor history (1/2)

Heat Transfer Reactor Experiment #3

ia/File:HTRE-3.jpg

The B-36 Aircraft Shield Test

[1]

13 ORNL Molten Salt Reactor Experiment (MSRE)

  • Operated from 1965 to 1969
  • 10 MWth
  • Used for SCALE MELCOR source term demonstration calculations Molten-salt reactors history (2/2)

MSRE

[ORNL-TM-0728]

MSRE Graphite Core Structure

[2]

14 Reactor

  • 10 MWth
  • Reactor consists of a graphite core structure (see photo on previous slide)
  • Fuel dissolved in the molten salt coolant fissions when it passes through the graphite core structure
  • Graphite provides moderation
  • 0.075 m3/s (1200 gpm) core flowrate
  • 635core inlet
  • 668core outlet
  • Near atmospheric pressure in the helium above the salt
  • INOR-8 nickel-based alloy vessel MSRE (1/5)

MSRE vessel

[ORNL-TM-0728]

15 Coolant salt circulation

  • Primary loop with pump and heat exchanger
  • Intermediate loop with pump and air-cooled radiator
  • No fuel in intermediate loop Air-cooled radiator rejects heat to the plant stack MSRE (2/5)

MSRE schematic

[ORNL-TM-0728]

MSRE primary heat exchanger

[3]

16 Reactor Cell acts as containment

  • Contains the reactor vessel, the primary circulating fuel loop, and most of the coolant salt loop
  • Circulating salt to air-cooled radiators located outside of the reactor cell MSRE (3/5)

MSRE Reactor Cell

[ORNL-TM-0728]

  • 95% N2
  • 0.875 bar absolute
  • 320 m3
  • Leak rate = 0.42 standard liters per hour at 0.875 bar (12.7 psia) (0.23 mm dia.)
  • Attached by a tunnel to the drain tank cell Reactor Cell Drain Tank Cell Reactor Building

17 Vapor-condensing system

  • Connects to the reactor cell via a 30 pipe
  • Normally isolated from the stack with 2 rupture disks MSRE (4/5)

MSRE vapor-condensing system

[ORNL-TM-0728]

30.5 cm (12) line with 1.38 bar (20 psig) rupture disk 10 cm (4) line with 1.03 bar (15 psig) rupture disk

  • Condensing tank with 34 m3 (1200 ft3) of water
  • Gas retention tank 93 m3 (3300 ft3)
  • 5 cm (2) line to the filters and the stack From the Reactor Cell

18 Off-gas filtration system

  • Large network that includes 6000 liter/day helium flow through the primary and secondary pump bowls
  • Pump bowl helium effluent connects to a series of holdup volumes (large volume &

low flow) inside and outside the reactor cell MSRE (5/5)

MSRE off-gas system

[ORNL-TM-0732]

  • 2 filter trains with 0.623 m3 (22 ft3) of charcoal One train typically isolated Auxiliary charcoal filter for reactor cell venting
  • 3x32 m2 (3x350 ft2) fiberglass roughing filters 90-95% efficiency for dust
  • 3x2.23 m2 (3x24 ft2) HEPA absolute filters with 99.7%

efficiency for 0.3 micron particles

  • Filtered flow merges with 9.9 m3/s (21,000 cfm) building HVAC out the plant stack for dilution Pump Bowl

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 Molten Salt Reactor Inventory, Decay Heat, Power, and Reactivity Methods and Results

20 Objectives:

  • Develop approach and models for SCALE analysis to obtain:
  • Radionuclide inventory
  • System decay heat
  • Power profiles
  • Reactivity coefficients Key differences to LWR analysis:

Continuous circulation of the fuel Consideration of both core and loop Nuclide removal in loop Approach:

  • Generate system fuel salt composition considering continuous circulation of the fuel salt and nuclide removal in the loop
  • Investigate location-dependent fuel salt inventory in the system
  • Evaluate neutronic characteristics at specific point in time NRC SCALE/MELCOR Non-LWR Demonstration Project SCALE MSRE core model

21 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*

  • The recently published NUREG/CR-7289 Nuclear Data Assessment for Advanced Reactors details the impact of the nuclear data library on non-LWR reactor physics calculations.

22 Basis for core model development: Zero-power first critical experiment with 235U from the OECD/NEA International Handbook of Reactor Physics Experiments [2]

MSRE Model Description

[1] R. C. Robertson (1965), MSRE Design and Operations Report Part I: Description of Reactor Design, ORNL-TM-0728, ORNL.

[2] M. Fratoni, et al. (2020), Molten Salt Reactor Experiment Benchmark Evaluation, DOE-UCB-8542, 16-10240, UC Berkeley, doi:10. 2172/1617123 MSRE reactor vessel [1]

Description Value Power 10 MWth (initial criticality) / 8 MWth (during operation)

Fuel/coolant LiF-BeF2-ZrF2-UF2 Enrichment 34.5 wt.% 235U Moderator Graphite Structure Nickel-based alloys Core volume 0.7 m3 System volume 2 m3 Heavy metal loading 0.233 tHM Loop transit time 25.2 seconds Nuclide removal Noble gases via Off-Gas System (OGS)

Noble metal plate-out at heat exchanger (HX)

Re-fueling Irregular re-fueling by capsules with HEU fuel salt Operating time

~375 equivalent full-power days with 235U fuel

23 SCALE analysis approach Core power/flux distribution

  • Predicts neutron flux and power profiles at point in time Time-dependent inventory
  • Considers core + loop +

off-gas + plating-out

  • Predicts system-average inventory over time Location-dependent inventory in loop
  • Considers power profile and off-gas
  • Predicts inventory in each region of the loop System-average inventory at point in time Power/flux profile System-average inventory at point in time Xe, Kr HX OGS Se, Nb, etc.

Core

+

loop

24 SCALE analysis approach Core power/flux distribution

  • Predicts neutron flux and power profiles at point in time Time-dependent inventory
  • Considers core + loop +

off-gas + plating-out

  • Predicts system-average inventory over time Location-dependent inventory in loop
  • Considers power profile and off-gas
  • Predicts inventory in each region of the loop System-average inventory at point in time Power/flux profile System-average inventory at point in time Xe, Kr HX OGS Se, Nb, etc.

Core

+

loop Sensitivity study:

Region-dependent nuclide inventory

25 SCALE analysis approach Core power/flux distribution

  • Predicts neutron flux and power profiles at point in time Time-dependent inventory
  • Considers core + loop +

off-gas + plating-out

  • Predicts system-average inventory over time Location-dependent inventory in loop
  • Considers power profile and off-gas
  • Predicts inventory in each region of the loop System-average inventory at point in time Power/flux profile System-average inventory at point in time Xe, Kr HX OGS Se, Nb, etc.

Core

+

loop Consistency assessment on removal rates

26 SCALE analysis approach Core power/flux distribution

  • Predicts neutron flux and power profiles at point in time Time-dependent inventory
  • Considers core + loop +

off-gas + plating-out

  • Predicts system-average inventory over time Location-dependent inventory in loop
  • Considers power profile and off-gas
  • Predicts inventory in each region of the loop System-average inventory at point in time Power/flux profile System-average inventory at point in time Xe, Kr HX OGS Se, Nb, etc.

Core

+

loop Power profiles and temperature-dependent reactivity coefficients Region-dependent inventory and decay heat for MELCOR System-average + OGS inventory and decay heat, and removal rates for MELCOR

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 MSRE full core model

28 TRITON-KENO model based on IRPhEP benchmark specifications MSRE full core model XY-cut through SCALE 3D model YZ-cut through SCALE 3D model graphite fuel fuel fuel fuel Cross section of graphite stringer

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 Time-dependent inventory

30 Goal: Generate system-average (fuel salt in core+loop) nuclide inventory at end of operation Model: TRITON-KENO core slice model

  • Representative spectral conditions through radial leakage and representative moderator-to-fuel ratio, while allowing shorter runtimes compared to full core
  • Depletion up to 375 days, the total operation time of MSRE with 235U fuel
  • Representation of system (core+loop) through adjusted power level:

Core power 8 MWth, total mass of 0.218 tHM in the system Specific power of 36.697 MW/tHM

  • Consideration of nuclide removal through TRITON-MSR 1,2 (next slide)

Time-dependent inventory - model development SCALE 2D slice model

[1] B. R. Betzler, al., Molten salt reactor fuel depletion tools in SCALE, Proc. Global/Top Fuel, Seattle, WA, September 22-27, 2019.

[2] P. J. V. Valdez, et al., Modeling Molten Salt Reactor Fission Product Removal with SCALE, ORNL/TM-2019/1418, 2020.

31 Nuclide removal via TRITON-MSR

  • Time-dependent removal of nuclides from one mixture into another
  • User-specified removal constant i,rem as used by ORIGEN to solve ODE:

Time-dependent inventory - nuclide removal bed stack plate-out core + loop tank/

OGS Production of nuclide i from decay and/or irradiation of nuclide j Source of nuclide i Loss rate of nuclide i due to decay, irradiation, or other means (flow)

+

32

  • Noble gas removal in the off-gas system:

Main experimental basis is the xenon poison fraction (ratio of absorption by 135Xe to absorption by 235U), reported as 0.3-0.4%

Noble gas removal fraction was set at 0.03 to match xenon poison fraction

  • Noble metal plating-out at the heat exchanger:

After operation, plated-out noble metals found, with 40% of noble metals plated out in heat exchanger, 50% on all other surfaces in the loop Noble metal plate-out removal rate determined from region-wise removal rates, as determined from mass transfer rate, surface area, and fuel salt volume Total removal rate calculated as sum of component-wise removal rates Time-dependent inventory - nuclide removal

33

- Depletion at low power level of 8 MWth, with flux level 1.881013 n/cm2-s

- No re-fueling in this depletion calculation

- At 375 days:

5.627% 235U consumed, 0.455% 238U consumed, 13.76 GWd/tHM burnup achieved Time-dependent inventory 0

50 100 150 200 250 300 350 400 10-11 10-10 10-9 10-8 10-7 10-6 10-5 Nuclide density [atoms/b-cm]

Days w/o Xe and Kr removal w/ Xe and Kr removal Xe Kr 1.11 1.12 1.13 1.14 1.15 1.16 1.17 k

k Amount removed after 375 days Noble gas (Xe + Kr) 0.170 kg / 30.6 L Insoluble metals (Mo + Tc + Ru + Rh + Pd + Ag + Sb) 0.611 kg Sometimes soluble metals (Se + Nb + Te) 0.057 kg Comparison of Xe and Kr nuclide densities with and without Xe/Kr removal

34 Decay heat after shutdown at 375 days System decay heat [% operating power]

Top contributors

35 Decay heat after shutdown at 375 days MSRE operating power: 8 MWth OGS HX

36

  • Side calculation with 235U feed through TRITON-MSR
  • Continuous feed rate of 1.49x10-3 g/s to yield approximately constant eigenvalue
  • Increasing 235U fuel concentration compensates for fission product buildup
  • Consider low burnup, and hardly any 239Pu buildup Demonstration of continuous feed / refueling Comparison of 235U nuclide densities and eigenvalue with and without 235U feed 0

50 100 150 200 250 300 350 400 7.8x 10- 5 8.1x 10- 5 8.4x 10- 5 8.7x 10- 5 9.0x 10- 5 9.3x 10- 5 9.6x 10- 5 k

w/o 235U feed w/ 235U feed Nuclide density [atoms/b-cm]

Days 235U 1.13 1.14 1.15 1.16 1.17 k

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 Core power/flux distribution

38 Used TRITON-KENO 3D full core model based on IRPhEP benchmark specifications as basis Analyzed 3D flux profiles and 3D fission rate via mesh tally capability informed discretization of core region Discretized model uses 34 axial and 8 radial zones Core power/flux distribution - model development Discretized model

39 Core power/flux distribution

- fission rate and flux Fission rate distributions Sample basket Lower end of graphite structure Upper end of graphite structure

40 Normalized radial power Core power/flux distribution - power

41 Core power/flux distribution - power Normalized axial power Top of graphite stringers Bottom of graphite stringers Different fuel-to-moderator ratios in upper and lower region cause small peaks in axial power

42 Impact of temperature distribution on the power profile Nominal case: 911 K in the fuel salt and graphite structure Temperature distribution from MELCOR: 910.5 -937.7 K for the fuel salt, 912.3-937.7 K for the graphite structure Core power/flux distribution - power Normalized axial power profile Normalized radial power profile

43 Determined reactivity coefficients by temperature/density perturbation:

  • Calculated reactivity at multiple temperature/density points
  • Fitted reactivity
  • Determined reactivity coefficient as derivative of fitted curve Core power/flux distribution - reactivity coefficients Component Fresh core 375 days eff [pcm]

704 +/- 14 697 +/- 22 Graphite temperature reactivity coefficient [pcm/K]

-5.13 +/- 0.05

-4.83 +/- 0.07 Fuel salt temperature and density reactivity coefficient [pcm/K]

-8.27 +/- 0.12

-8.28 +/- 0.12

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 Location-dependent inventory in loop

45 Developed ORIGEN model to predict nuclide inventory in each region of the loop at ~375 days

  • Divided MSRE system into 9 general regions, with the core region subdivided into 30 axial zones
  • Used fuel salt composition from 2D TRITON-MSR calculation at 375 days as the start
  • Developed chain of ORIGEN inputs that use residence time and flux of the fuel salt in each region and removes noble gases (Kr, Xe) in off-gas system
  • 1 ORIGEN input corresponds to the salt traveling 1 time through the whole loop Location dependent inventory - model development Regions in MSRE system for ORIGEN model 30 axial zones

46 As fuel salt travels the loop Long-lived*

nuclides will slowly accumulate/be removed Short-lived*

nuclides will oscillate around an equilibrium Equilibrium established after a

few loops (resulting in inventory at just a few minutes after 375 days)

Location dependent inventory - model development

  • relative to the loop transit time (~25 s for MSRE)

Regions in MSRE system for ORIGEN model 30 axial zones

47

  • Observed constant densities of long-lived nuclides for several loops
  • Observed convergence of short-lived nuclides after ~6 loops Location dependent inventory analysis example Example: Short-lived nuclide (I-137, t1/2=24.5s) as a function of time at the bottom of the core I-137 at the bottom of the core 0

50 100 150 200 250 300 1.16x 10- 13 1.20x 10- 13 1.24x 10- 13 1.28x 10- 13 1.32x 10- 13 Nuclide Density [atoms/b-cm]

Time [s]

48

  • Compared short-lived nuclide densities between different regions
  • Found that inventory/decay heat does not significantly differ between regions when summed up into element classes due to short loop transit time in MSRE Location dependent inventory analysis example Example: Short-lived nuclide (I-137, t1/2=24.5s) as a function of location in the loop 1.

Core 2.

Upper head 3.

Piping to Pump 4.

Pump/OGS 5.

Piping to HX 6.

HX 7.

Piping to RX 8.

Inlet 9.

Lower head 0

5 10 15 20 25 1.1x 10- 13 1.2x 10- 13 1.3x 10- 13 1.4x 10- 13 1.5x 10- 13 1.6x 10- 13 Nuclide density [atoms/b-cm]

Time [s]

1D loop System average Core (1) 2 3

4 5

9 8

7 6

49 Delayed neutrons are important for reactivity control Fission products that emit delayed neutrons are called delayed neutron precursors (DNP)

In flowing fuel systems, delayed neutrons may be born outside of the core, commonly called DNP drift For example MSRE eff ~ 700 pcm without drift eff decreases as flow speed (drift) increases A DNP drift model has not yet been incorporated in this work Sensitivity studies show using detailed axial-dependent nuclide density versus system-average has negligible effect on the core power shape Delayed neutron precursor drift 10- 15 10- 14 10- 13 0.0 40.0 80.0 120.0 160.0 200.0 240.0 Upper head Height [cm]

Nuclide density [atoms/b-cm]

87Br 137I 88Br 93Rb 138I 94Rb 89Br 139I 90Br Lower head Core Selected delayed neutron precursors calculated by ORIGEN

50

  • Two approaches will be pursued MELCOR DNP drift model based on standard 6-group delayed neutron precursors New higher-fidelity model in MELCOR based on explicit delayed neutron precursor nuclides, as available through ORIGEN Delayed neutron precursor drift (cont.)

10- 15 10- 14 10- 13 0.0 40.0 80.0 120.0 160.0 200.0 240.0 Upper head Height [cm]

Nuclide density [atoms/b-cm]

87Br 137I 88Br 93Rb 138I 94Rb 89Br 139I 90Br Lower head Core Selected delayed neutron precursors calculated by ORIGEN

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 Summary of SCALE methods and results

52 SCALEs capabilities were demonstrated:

  • 3D modeling with TRITON-KENO for time snapshots of power profiles and reactivity coefficients
  • TRITON-MSR for time-dependent system-average inventory considering noble gas and noble metal removal through off-gas system and plating out, respectively
  • ORIGEN for region-dependent inventory considering noble gas removal Planned enhancements:
  • TRITON-MSR with continuous feed
  • Tracking of removed nuclides in ORIGEN
  • Integration of ORIGEN into MELCOR SCALE MSR Summary

MELCOR Molten Salt Reactor Models

54 Molten Salt Reactor modeling in MELCOR

  • Accident progression
  • Source term Molten Salt Reactor Modeling Contain Radionuclide
species, transport, and retention Cool Generalized EOS, CVH, HS Control Fluid point
kinetics, transmutation

55 MELCOR remains a general purpose, multi-physics code to model integral plant response under accident conditions Serves as an effective foundation to support NRC readiness to license advanced nuclear energy technologies Fluid fuel thermal hydraulics Leverage existing thermal hydraulics modeling in MELCOR Utilize fundamental two-phase thermal hydraulic equations Introduce new thermo-physical properties and phase diagram of fluid specific to FLiBe Generalized EOS - Equations of state for multiple working fluids are presently available in MELCOR including water, sodium, and FLiBe Thermal hydraulics - CVH/FL Model Packages Control Volume Hydrodynamics (CVH) package defines control volumes (CV)

Flow path modeling package defines flow paths (FL)

Heat Transfer - HS/CVH/COR Packages The HS package defines heat structures (HS) that model radiative and conductive heat losses CVH package manages convective heat losses Modeling MSR Accidents with MELCOR -

Hydrodynamics and Heat Transport CV WW CV XX CV YY CV ZZ FL WY FL WX FL XZ FL YZ HS ZZ H

S X

Z

56 Fluid fuel point kinetics enables simplified, but appropriate treatment of neutronic transients Fuel point kinetics - derived from standard point kinetic equations and solved similarly Range of feedback models available for flexible modeling of transients User-specified external input Other implementations in the code (e.g., Doppler, fuel and moderator density) generally not used for MSR applications because they were derived for other types of reactor cores Flow reactivity feedback effects integrated into the equation set Control volume fluid core with power distribution Neutronics model provides power in core-region, distribution of precursor radionuclides in the core and around the loop Radionuclides advected with the flowing salt contribute to decay heat in different regions of the reactor Fission product transmutation enhancement Coupling with SCALE/Oak Ridge Isotope GENerator (ORIGEN) ongoing Modeling MSR Accidents with MELCOR -

Reactivity Control

57 Molten salt serves as a potential means of fission product retention Generalized Radionuclide Species

  • Users can redefine/add RN classes RN classes exhibit similar transport and retention behavior Approach taken for molten salt systems - unique fission product chemistry relative to water-moderated systems See Slide 75 for example grouping chosen for MSRs Fluid fuel radionuclide transport
  • Generalized Radionuclide Transport and Retention (GRTR) modeling framework
  • Molten salt chemistry and physics pertaining to radionuclide transport
  • GRTR for MSRs but generalized and applicable to other systems (e.g., liquid metal)

Modeling MSR Accidents with MELCOR - Fission Product Transport and Release

58 Generic working fluid EOS capability facilitates FLiBe as hydrodynamic material

  • MELCOR employs fluid property files - INL fusion safety program
  • Chens soft sphere model used for FLiBe (INL/EXT-17-44148)
  • Property database from ORNL data (ORNL-TM-2316)
  • Verified MELCOR EOS library and properties for FLiBe Initial validation activity against ORNL MSRE FLiBe Equation of State

59 Freezing of molten salt an important consideration for a range of accident conditions Address fluids that freeze in an accident such as a salt spill Freezing in cooling systems (e.g., DRACS)

Adding capabilities to explicitly treat freezing of fluids Currently an approximation is used to handle conditions where fluids reach temperatures at or below their freezing point Generalized capability for other fluids (e.g., sodium)

FLiBe Equation of State - Implications of Salt Freezing

60 Fluid fuel core defined within the graphite stringers

  • The fluid volume within the graphite stringers comprise the active Core
  • Loop volumes comprise a portion of the primary fuel flow loop OUTSIDE the active core
  • Allows specification of the axial and radial power distribution from SCALE Feedbacks and power governed by flowing fluid fuel point reactor kinetics model Fission power generation in core and loop control volumes
  • Fission power and feedbacks are calculated for the core volumes
  • No fission power energy generation in loop volumes
  • Decay heat (due to radionuclide class mass carried in pool) for both volume types
  • Graphite heating due to neutron absorption
  • Provisions for shutdown in a spill accident Fluid Core and Power Distribution

61 Fluid Fuel Neutronic Transients - Modified Point Kinetics A

B C

D E

A - In-Vessel DNP gain by fission B - In-Vessel DNP loss by decay and flow C - In-Vessel DNP gain by Ex-Vessel DNP flow D - Ex-Vessel DNP gain by In-Vessel DNP flow E - Ex-Vessel DNP loss by decay, flow Fission inside core Neutrons generated and moderated DNPs generated DNPs that do not decay in core-region flow into loop Decay in loop or advect back into core-region

    • DNP = Delayed Neutron Precursor

62 Fluid Fuel Point Kinetics - Initial Validation MELCOR non-LWR validation is leveraging available data Validation basis will continually expand with evolution of tests and deployments Initial validation has been performed against zero-power MSRE pump flow coast-down test

63 Track where fission products are and how much is released from liquid to atmosphere Characterizes evolution of fission products between different physico-chemical forms Fission product evolution from a liquid pool to an atmosphere Influenced by solubility and vapor pressure Insoluble fission product deposition on structures GRTR mass transport modeling characterizes Concentration of radionuclide forms Concentration gradients between radionuclide forms Resistance to mass transfer between radionuclide forms using standard correlation-based interfacial mass transport theory GRTR - Generalized Radionuclide Transport and Retention

64 GRTR and Integral MELCOR Simulations For Each Timestep Inputs to GRTR Model Radionuclide mass in (or released to) liquid pool Chemical speciation Pressure in hydrodynamic volume Temperature in regions of hydrodynamic volume (e.g., liquid and atmosphere)

Advective flows of liquid and atmosphere between hydrodynamic volumes GRTR Physico-Chemical Transport Dynamics Soluble radionuclide form mass Colloidal radionuclide form mass Deposited radionuclide form mass Gaseous radionuclide mass Advective and Fission/Transmutation Dynamics Advection of radionuclides in liquid pool or atmosphere Decay of radionuclides in hydrodynamic control volume Coupling with ORIGEN

65 Evolution of fission products from molten salts primarily focused on vaporization Provides ability to perform best estimate evaluations of release from molten salts Demonstration calculations have focused on direct comparison to MSRE for the maximum hypothetical accident Exercise of model will be performed next year Mass transfer interfaces Liquid-gas atmosphere interfaces Liquid-solid structure interfaces Gas atmosphere-solid structure interfaces Model allows new interfaces to be defined Sparging gas flows (i.e., helium gas injection) will result in fission products entrained in the gas bubble formed by injection Jet breakup when contaminated fluids are released into a gas atmosphere (e.g., due to a pipe break)

GRTR - Range of Mass Transport Processes

66

=

Illustrating Components of Vaporization Mass Transfer Mass transfer coefficient captures effectiveness of species diffusion into atmosphere from liquid-gas interface as well as convective flows carrying vapor away from interface Example CsF vapor pressure - subject to change as thermochemistry evolves

67 Focus of modeling efforts evolving based on insights from current demonstration calculations For salt spills, MELCOR GRTR model predicts very small vaporization releases of CsF, CsOH and CsI from salt Relatively low temperature molten salt temperature leads to a very low vaporization (<<10-6)

Contribution of the vaporization term in a spill scenario is negligible Ongoing model development utilizes flexibility to explore different ways to characterize other release mechanisms

  • Jet breakup and splashing models
  • Aerosol release from bubble bursting Evolution of GRTR Modeling

Molten Salt Reactor Plant Model and Source Term Analysis

69 MELCOR nodalization - core and reactor vessel Vessel nodalization

  • Assumes azimuthal symmetry
  • The graphite core structure is subdivided into 10 axial levels and 5 radial rings Next slide shows mapping from SCALE
  • Molten fuel salt enters through an annular distributor (cv-100) that directs the flow into the annular downcomer (cv-105) and the core inlet plenum (cv-110)
  • The core is formed by graphite stringers that include flow channels
  • The molten fuel salt flows through the stringers (CV-210 through CV-259), where the fuel fissions Core region v

v v

v v

v v

v v

v v

v v

v v

v v

v v

v 219 218 217 216 215 214 213 212 211 210 229 228 227 226 225 224 223 222 221 220 239 238 237 236 235 234 233 232 231 230 249 248 247 246 245 224 243 242 241 240 259 258 257 256 255 254 253 252 251 250 110 - Inlet Plenum v

105 Annulus Downcomer 130 - Outlet Plenum 100

70 MELCOR core region mapping to SCALE 600 stringers ORNL Radial Zone (r) 1 & 2 3

4 5

6 & 7 MELCOR Radial Zone (r) 1 2

3 4

5 Percent 3.5%

19.3%

35.3%

36.7%

5.2%

1 2 3

4 5

6 7

38.55cm 159.24cm 35.93cm 1

34 3-32 2

33 MELCOR axial mapping is 3 SCALE levels per 1 MELCOR level MELCOR radial mapping to SCALE

71 150 306 398 403 400 MELCOR nodalization - primary recirculation loop 306 305 310 300 330 320 321 401 402 Total core power Fission = 8.8 MW Graphite heating = 0.7 MW Decay heat = 0.4 MW Flows Primary loop = 1200 gpm Intermediate loop = 850 gpm Helium off-gas flows Pump shaft = 1279 l/d Pump bowl = 3456 l/d Overflow tank = 1279 l/d Recirculation flows Pump bowl spray = 50 gpm Pump shaft = 15 gpm

72 MELCOR nodalization - reactor cell, condensing tank, and reactor building Leakages

  • Reactor cell = 0.42 scfh at 12.7 psia
  • Reactor bldg = 10% per day at 0.25 psig Reactor Building CV-520 FL-525 FL-520 HVAC supply HVAC exhaust FL-599 Bldg leakage FL-515 Reactor cell leakage CV-530 FL-550 Vacuum brkr Water FL-545 Condensing tank CV-535 Gas retention tank CV-540 FL-555 FL-560 To filters & stack FL-525 - Vacuum pump Rupture disks FL-535 = 15 psig (4 line)

FL-540 = 20 psig (12 line)

To the stack Closed valves CV-525 30 vent line FL-535 FL-540 150 Reactor Cell Drain Tank Room CV-510 CV-515 FL-510 CV-399 Pump furnace Bldg leakage FL-598

73 MELCOR nodalization - offgas system From the pump bowl cv-600 cv-605 cv-601 Water-cooled flow Charcoal beds cv-610 cv-615 Water-cooled flow cv-635 Aux. Charcoal beds cv-620 cv-625 Roughing filter Absolute filter cv-699 Plant stack Filter pit Building HVAC

74 MELCOR model inputs (1/2)

Equilibrium inventory and decay heat by region from SCALE Radial and axial power profiles from SCALE Reactivity and Xe feedbacks from SCALE Radionuclide distribution from SCALE Collaborative redefinition of radionuclide classes with ORNL

  • Re-grouping from LWR definition based on solubility estimations from MSRE empirical experience and suggestions by Britt (ORNL)
  • Noble metals isolated into two groups (next slide)

Antoine coefficient estimates for few species (Cs, CsF)

[Phillip Britt, Future Research Directions, DOE-NE Molten Salt Chemistry Workshop, April 10-12, 2017, ORNL]

75 MELCOR model inputs (2/2)

MELCOR Elemental Grouping Xe : He, Ne, Ar, Kr, Xe, Rn, H, N Cs : Li, Na, K, Rb, Cs, Fr, Cu Ba : Be, Mg, Ca, Sr, Ba, Ra, Es I : F, Cl, Br, I, At S : S, Po Re : Re, Os, Ir, Pt, Au, Ni V : V, Cr, Fe, Co, M, Ta, W Mo : Mo, Tc, Ru, Rh, Pd, Ag, Ge, As, Sn, Sb Nb : Nb, Zn, Cd, Se, Te Ce : Ti, Zr, Hf, Ce, Th, Pa, Np, Pu, C La : Al, Sc, Y, La, Ac, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Am, Cm, Bk, Cf U : U Cd : Hg, Ga, In Ag : Pb, Tl, Bi B : B, Si, P

[Phillip Britt, Future Research Directions, DOE-NE Molten Salt Chemistry Workshop, April 10-12, 2017, ORNL]

Mo class assumed to be insoluble

76 Spill of molten salt into the reactor cell (containment)

  • Full reactor spill - maximum credible accident in the MSRE safety analysis Spill onto the floor without coincident water leak (MCA1-MCA5)

Spill with coincident water leak (MCA6-MCA9)

Exploratory radionuclide source term due to limited information from the molten salt thermophysical databases

  • ORNL-TM-0732 MSRE safety analysis source term Integral calculation with aerosol physics
  • GRTR vaporization model without splashing Cs, CsI, and Xe releases Sensitivities
  • HVAC operating or off
  • Auxiliary filter operation
  • Aerosol size Scenario

77 Case Aerosol size Stack Fans Aux. Filters Water Spill MCA1 1 µm Yes No No MCA2 10 µm Yes No No MCA3 1 µm No No No MCA4 1 µm Yes Yes No MCA5 1 µm No Yes No MCA6 1 µm Yes No Yes MCA7 1 µm Yes Yes Yes MCA8 1 µm No Yes Yes MCA9 1 µm No No Yes Salt spill cases Walk-through MCA1 (base case)

Spill creates aerosols with 1 µm mass median diameter (MMD) with a 1.5 geometric standard deviation (GSD)

The HVAC remains running and ventilating the reactor building The auxiliary filters are not used to filter the reactor cell There is no coincident water spill onto the molten salt

78 MCA1 salt spill base case - Primary System Response

79 MCA1 salt spill base case - Reactor Cell Response

80 0

100 200 300 400 500 600 700 800 900 1000 0

5 10 15 20 25 0

10 20 30 40 50 60 Temperature (F)

Pressure (psia)

Time (min)

Pressure Temperature Normal pressure (12.7 psia) 500 600 700 800 900 1000 1100 1200 1300 1400 0

1000 2000 3000 4000 5000 6000 7000 8000 9000 0

10 20 30 40 50 60 Temperature (F)

Mass (lbm)

Time (min)

Mass Temperature Reactor cell pressure and temperature

  • The primary loop salt inventory spills to the reactor cell in 10 minutes Temperature of the molten salt is relatively constant with a slow cooling trend
  • There is an immediate pressurization of the gas space from subatmospheric to ~20 psia Heating due to hot molten salt (~1100F)

Heating due to the released radionuclides

  • Reactor cell gas temperature initially rises to over 900F and then slowly cools MCA1 reactor cell thermal-hydraulic response Mass of molten salt spilled and its temperature

81 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03

-400 0

400 800 1200 1600 2000 2400 2800 3200 3600 Pressure (psi)

Temperature (°F)

I2 Cs metal CsOH CsI Ce 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1

10 100 1000 Fraction of the initial inventory (-)

Time (sec)

Xe Cs Iodine (gas)

Iodine (aerosol)

Ce (Pu)

Xe (noble gases)

All others

  • Airborne release assumptions

[ORNL-TM-0732]

100% of the noble gases 10% of the iodine 10% of all other volatile and non-volatile radionuclides MCA from MSRE safety analysis

  • Radionuclides phase (aerosol or gas) depends on temperature and chemical form Preliminary analysis used LWR view of chemical forms Some gaseous iodine (5%)

Cesium and iodine combine (CsI)

MELCOR allows exploration of various chemical forms for the MSR MCA1 reactor cell radionuclide releases Radionuclide airborne release Vapor pressure in the atmosphere

82 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Reactor bldg Environment Reactor cell Environment Reactor bldg 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Reactor bldg Environment Reactor cell Environment Reactor bldg Gaseous releases (xenon and iodine gas) respond similarly

  • Most of the gases retained in the reactor cell
  • Reactor cell slowly leaks to the reactor building The reactor building HVAC is operating in MCA1, which exhausts gases from the reactor building through the absolute filters to the plant stack
  • 0.2% of the xenon reaches the environment
  • 0.02% of the gaseous iodine reaches the environment MCA1 gaseous radionuclide distributions Xe distribution Iodine gas distribution From the pump bowl cv-600 cv-605 cv-601 Water-cooled flow Charcoal beds cv-610 cv-615 Water-cooled flow cv-635 Aux. Charcoal beds cv-620 cv-625 Roughing filter Absolute filter cv-699 Plant stack Filter pit Building HVAC

83 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Reactor cell airborne Reactor bldg Filters Environment Reactor cell Environment Reactor bldg Filters 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Reactor cell airborne Reactor bldg Filters Environment Reactor cell Environment Reactor bldg Filters Release fractions of radionuclides that form aerosols in the reactor building

  • CsOH and CsI illustrate radionuclide chemical forms that are primarily vapor in the reactor cell with limited settling but aerosols after leakage to the reactor building
  • Ce distribution is typical of less volatile radionuclides that settle over time in the reactor cell and are captured by the absolute filters MCA1 aerosol radionuclide distributions CsOH distribution CsI distribution 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Reactor cell airborne Reactor bldg Filters Environment Reactor cell Environment

(<10-8)

Reactor bldg Filters Ce distribution

84 Salt spill with water base case Walk-through MCA6 Spill creates aerosols with 1 µm mass median diameter (MMD) with a 1.5 geometric standard deviation (GSD)

The HVAC remains running and ventilating the reactor building The auxiliary filters are not used to filter the reactor cell Water spill onto the molten salt Case Aerosol size Stack Fans Aux. Filters Water Spill MCA1 1 µm Yes No No MCA2 10 µm Yes No No MCA3 1 µm No No No MCA4 1 µm Yes Yes No MCA5 1 µm No Yes No MCA6 1 µm Yes No Yes MCA7 1 µm Yes Yes Yes MCA8 1 µm No Yes Yes MCA9 1 µm No No Yes Equilibration of all the fuel salt with the cell atmosphere and just enough water to form the maximum amount of saturated steam would result in the maximum pressure in the secondary container.

With no relief device, pressures as high as 110 psig could result.

[ORNL-TM-0732]

[ORNL-TM-0732]

85

  • Molten salt is assumed to mix with coincidentally spilled water Rapid pressurization of the reactor cell as it fills with steam
  • Reactor cell pressure rises to 46 psia 15 psi rupture disk opens at 41 sec 20 psi rupture disk opens at 115 sec
  • Reactor cell temperature initially rises to 330F but falls after the 20 psi rupture disk opens MCA6 reactor cell thermal-hydraulic response Reactor cell and gas retention tank pressure and temperature MSRE vapor-condensing system

[ORNL-TM-0728]

0 100 200 300 400 500 0

5 10 15 20 25 30 35 40 45 50 0

100 200 300 400 500 600 Temperature (F)

Pressure (psia)

Time (sec)

Reactor cell pressure Gas retention tank pressure Reactor cell temperature Gas retention tank temperature 15 psi rupture disk 20 psi rupture disk

86 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Cond & gas retention tanks Reactor bldg Environment 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Cond and gas retention tanks Reactor bldg Environment Same MCA airborne releases into reactor cell

  • Release assumptions 100% of the noble gases 10% of the iodine 10% of all other volatile and non-volatile radionuclides
  • Strong flows to the condensing and gas retention tanks capture most of the radionuclides released from the spilled salt Condensing tank retains most of the aerosols and the gas retention tank captures any radionuclides that pass through the pool All noble gases and most of the gaseous iodine passes through the condensing pool MCA6 gaseous radionuclide distributions Xe distribution Iodine gas distribution

87 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Reactor cell airborne Condensing tank Reactor bldg Filters Environment 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Reactor cell airborne Condensing tank Reactor bldg Filters Environment 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1

10 100 1000 10000 100000 Fraction of the initial inventory (-)

Time (sec)

Reactor cell Reactor cell airborne Condensing tank Reactor bldg Filters Environment Most of the aerosol releases are retained in the condensing tank

  • CsOH and CsI form aerosols in a water spill accident and behave similarly to the cerium aerosols The large steam source contributes to aerosol agglomeration and more rapid settling in the reactor cell than the dry case MCA6 aerosol radionuclide distributions CsOH distribution Ce distribution CsI distribution

88 The xenon release to the environment spanned many orders of magnitude depending on scenario assumptions

  • Lowest releases with no HVAC and no Aux filter flow
  • Auxiliary filter operation increases the release of xenon to the environment while it provides filtering of airborne aerosols Overall insights (1/4)

All results of xenon release to the environment 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 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Release Fraction (-)

Time (hr)

Xe release to the environment MCA1 MCA2 MCA3 MCA4 MCA5 MCA6 MCA7 MCA8 MCA9 Note: Results assume no xenon retention in the charcoal filters.

Case Aerosol size Stack Fans Aux. Filters Water Spill MCA1 1 µm Yes No No MCA2 10 µm Yes No No MCA3 1 µm No No No MCA4 1 µm Yes Yes No MCA5 1 µm No Yes No MCA6 1 µm Yes No Yes MCA7 1 µm Yes Yes Yes MCA8 1 µm No Yes Yes MCA9 1 µm No No Yes No HVAC + no Aux filter cases HVAC + Aux filter Aux filter cases

89 The aerosol releases to the environment were small due to:

  • Gravitational settling in the reactor cell (all cases),

the reactor building, filter pit, and stack (without HVAC flow)

  • Capture in the filter
  • Capture in the condensing tank in the water spill cases Overall insights (2/4) 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 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce airborne in the Reactor Cell MCA1 MCA2 MCA3 MCA4 MCA5 MCA6 MCA7 MCA8 MCA9 All results of cerium airborne fraction in the reactor cell Water spill cases No water spill cases 1.E-03 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce in the Condensing Tank MCA1 MCA2 MCA3 MCA4 MCA5 MCA6 MCA7 MCA8 MCA9 No spill cases = 0%

Water spill cases = ~65%

All results of cerium capture in the condensing tank All results of cerium release to the environment 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 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce release to the environment MCA1 MCA2 MCA3 MCA4 MCA5 MCA6 MCA7 MCA8 MCA9

90 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 1.E-02 1.E-01 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce release to the environment MCA6 MCA7 MCA8 MCA9 Due to the high temperatures in the reactor cell in the cases without a water spill (~900F), the two chemical compounds of cesium were primary in a vapor form

  • Any released CsI and CsOH subsequently condensed in the reactor building and the offgas system to form aerosols
  • CsOH and CsI remained airborne in the reactor cell versus cerium, which was always an aerosol This led to higher cesium environmental releases than radionuclides that were aerosols in the reactor cell (e.g., cerium)

Overall insights (3/4)

Spill results of cesium release to the environment 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 1.E-02 1.E-01 0

6 12 18 24 Release Fraction (-)

Time (hr)

Cs release to the environment MCA1 MCA2 MCA3 MCA4 MCA5 No spill results of cesium release to the environment 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03

-400 0

400 800 1200 1600 2000 2400 2800 3200 3600 Pressure (psi)

Temperature (°F)

I2 Cs metal CsOH CsI Ce

91 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce in the Reactor Building MCA1 MCA2 MCA3 MCA4 MCA5 MCA6 MCA7 MCA8 MCA9 The aerosol mass in the reactor building also spanned many orders of magnitude depending on scenario assumptions

  • Lowest amounts occurred when aerosols were captured in the condensing tank and any leaked aerosols were filtered via the reactor building HVAC flow
  • The highest amounts occurred when there was no water spill and no HVAC
  • Aerosols leaked into the reactor building without HVAC operation primarily settled (flat line)

Led to a small amount of leakage to the environment

  • Finally, the HVAC swept long-term releases into the reactor building in the no spill cases Overall insights (4/4)

All results of cerium in the reactor building No water spill & no HVAC Water spill cases & HVAC No water spill & HVAC Water spill & no HVAC

92 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 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Fraction of initial inventory (-)

Time (hr)

Ce in the Reactor Building and Environment 100X reactor cell leak 10X reactor cell leak 1X reactor cell leak 100X reactor cell leak 10X reactor cell leak 1X reactor cell leak 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 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Fraction of initial inventory (-)

Time (hr)

Xe in the Reactor Building and Environment 100X reactor cell leak 10X reactor cell leak 1X reactor cell leak 100X reactor cell leak 10X reactor cell leak 1X reactor cell leak

  • Green line shows impact of reactor cell leakage to the reactor building for the MCA3 scenario Gaseous Xe leak to the reactor building continues while the Ce aerosol leakage stops early due to aerosol settling
  • Blue line shows impact of reactor cell leakage on environmental releases for the MCA3 scenario The leakage to the reactor cell has an approximately linear impact on the reactor cell leak rate versus a slightly larger than linear effect on the environment leakage
  • The impacts are expected to be smaller with HVAC operation Sensitivity to increased reactor cell leakage Green - Xe in the reactor building versus reactor cell leak rate Blue - Xe in the environment for 1X containment leak rate Case Aerosol size Stack Fans Aux. Filters Water Spill MCA1 1 µm Yes No No MCA2 10 µm Yes No No MCA3 1 µm No No No MCA4 1 µm Yes Yes No MCA5 1 µm No Yes No Cases without a water spill Green - Ce in the reactor building versus reactor cell leak rate Blue - Ce in the environment for 1X containment leak rate

93 1.E-08 1.E-07 1.E-06 0

6 12 18 24 Fraction of initial inventory (-)

Time (hr)

Xe in the Environment 0 mph 100X RB leak 0 mph 10X RB leak 0 mph 1X RB leak 5 mph 100X RB leak 5 mph 10X RB leak 5 mph 1X RB leak 10 mph 100X RB leak 10 mph 10X RB leak 10 mph 1X RB leak

  • The reactor building surrounds the reactor cell and provides the final barrier for leakage when the filters are not operating
  • Impact of reactor building leakage as a function of wind speed shows a very small impact on the release to the environment for the MCA3 scenario Similar to increased building leakage, a higher wind speed increases the building infiltration and exfiltration rate The impact is slightly larger for gas leakage (i.e., aerosols also settle)
  • The nominal (1X) building leakage is very low Only 10% per day at 0.25 psig, 0.002 in2 The very large building (480,000 ft3) has no appreciable pressurization (i.e., <<0.25 psig)

Sensitivity to increased reactor building leakage Xe in the environment versus RB leakage and wind speed Case Aerosol size Stack Fans Aux. Filters Water Spill MCA1 1 µm Yes No No MCA2 10 µm Yes No No MCA3 1 µm No No No MCA4 1 µm Yes Yes No MCA5 1 µm No Yes No Cases without a water spill 1.E-12 1.E-11 1.E-10 1.E-09 0

6 12 18 24 Fraction of initial inventory (-)

Time (hr)

Ce in the Environment 0 mph 100X RB leak 0 mph 10X RB leak 0 mph 1X RB leak 5 mph 100X RB leak 5 mph 10X RB leak 5 mph 1X RB leak 10 mph 100X RB leak 10 mph 10X RB leak 10 mph 1X RB leak Ce in the environment versus RB leakage and wind speed

Summary

95 Conclusions Demonstrated use of SCALE and MELCOR for MSRE 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

98 Nuclide removal from fuel salt in core+loop:

  • Plating-out of noble metals (Se, Nb, Mo, Tc, Ru, etc.)

at heat exchanger

  • Removal of halogens (I, Br) from plated-out material
  • Removal of noble gases from plated-out materials
  • Removal of noble gases (Xe, Kr) from fuel into off-gas system
  • Removal of gas into charcoal bed
  • Removal of gas into stack Time-dependent inventory - nuclide removal bed stack plate-out at heat exchanger core + loop tank Figure modified from: R. C. Robertson (1965), MSRE Design and Operations Report Part I: Description of Reactor Design, ORNL-TM-0728, ORNL.

99 Core power/flux distribution - flux Thermal flux

(< 0.625 eV)

Fast flux

(> 0.625 eV)

100 Region-wise data

MELCOR for Accident Progression and Source Term Analysis

102 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).

103 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

104 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 preparedness
  • Emergency Planning Zone Analysis Design/Operational Support
  • Design analysis scoping calculations
  • Training simulators Fusion
  • Neutron beam injectors
  • 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

105 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

106 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 Asian MELCOR User Group (AMUG) Meeting - Fall/Asia

107 Common Phenomenology

108 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

110 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)

111 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]

112 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

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

Cs vapor pressures in GRTR calculations

115 Enhancement to MELCOR radionuclide transport modeling Incorporate unique chemistry of fission products in new fluids potentially mitigating release to atmospheres of reactor vessel, off-gas systems, and confinement/containment Retention in fluids influenced by physico-chemical form of fission products - strong influence of thermochemistry Is the fission product compound soluble?

Is the fission product compound insoluble (i.e., colloidal)?

Is the fission product compound a gaseous vapor?

Has the fission product compound deposited on a structural surface?

Is the fission product located at a liquid-atmosphere interface?

Interface between liquid pool and overlying gas atmosphere Interface between liquid and gas bubbles (e.g., generated by sparging helium gas)

Introduce new physico-chemical forms that supplement existing MELCOR representation of distinct radionuclide classes Soluble fission products Insoluble/colloidal fission products Deposited fission products Gaseous fission products Each tracked form is identified with either a liquid pool, an atmosphere, or deposited on a structure GRTR - Generalized Radionuclide Transport and Retention

116 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 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Release Fraction (-)

Time (hr)

Xe release to the environment MCA1 MCA2 MCA3 MCA4 MCA5 Gaseous xenon release to the environment without a water spill

  • MCA4 had the highest release due to auxiliary filter venting of the reactor cell after 1-hr and enhanced leakage due the HVAC flow through the reactor building
  • MCA3 had the lowest release with no HVAC flow in the reactor building (stack fans) and no auxiliary flow
  • MCA1 and MCA2 were identical because xenon is not an aerosol
  • MCA5 did not have enhanced releases due to HVAC venting the reactor building leaks but did include the auxiliary filter flow after 1-hr Results of the sensitivity studies Cases without a water Spill Case Aerosol size Stack Fans Aux. Filters Water Spill MCA1 1 µm Yes No No MCA2 10 µm Yes No No MCA3 1 µm No No No MCA4 1 µm Yes Yes No MCA5 1 µm No Yes No Cases without a water spill Note: Results assume no xenon retention in the charcoal filters.

117 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 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Release Fraction (-)

Time (hr)

Xe release to the environment MCA6 MCA7 MCA8 MCA9 Gaseous xenon release to the environment with a water spill

  • MCA7 had the highest release due to auxiliary filter venting of the reactor cell after 1-hr and enhanced leakage due the HVAC flow through the reactor building
  • MCA7 had a lower release than the corresponding dry case due to xenon capture in the gas retention tank
  • MCA9 had the lowest release due to no HVAC flow in the reactor building (stack fans) and no auxiliary filter flow
  • Leaks into the reactor building from MCA6 were vented to the environment due to the HVAC operation
  • MCA8 did not have enhanced releases due to HVAC venting any reactor building leaks but did include venting to the environment from the auxiliary filter flow after 1-hr Results of the sensitivity studies Cases with a water Spill Cases with a water spill Case Aerosol size Stack Fans Aux. Filters Water Spill MCA6 1 µm Yes No Yes MCA7 1 µm Yes Yes Yes MCA8 1 µm No Yes Yes MCA9 1 µm No No Yes Note: Results assume no xenon retention in the charcoal filters.

118 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 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce release to the environment MCA1 MCA2 MCA3 MCA4 MCA5 The cerium aerosol release to the environment without a water spill were very low

  • MCA3, MCA4, and MCA5 releases to the environment were approximately the same and larger than MCA1 MCA4 and MCA5 included continuous venting of very small aerosols from the reactor cell through the auxiliary filter MCA3 results show impact of nominal leakage from the reactor building (i.e., no filtration)

MCA1 included filtration of the reactor building but no auxiliary filter flow

  • MCA2 had larger aerosols, which settled faster and the smallest amount released to the environment Results of the sensitivity studies Cases without a water Spill Case Aerosol size Stack Fans Aux. Filters Water Spill MCA1 1 µm Yes No No MCA2 10 µm Yes No No MCA3 1 µm No No No MCA4 1 µm Yes Yes No MCA5 1 µm No Yes No Cases without a water spill Note: The capture efficiency of the absolute filters for aerosols below <0.3 µm was assumed to be zero.

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 1.E-02 1.E-01 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce settling and filter behavior MCA1 airborne in the reactor cell MCA1 captured by filters MCA1 environment MCA2 airborne in the reactor cell MCA2 captured by filters MCA2 environment Environment Filters Airborne in the reactor cell 1 versus 10 µm aerosol behavior

119 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 1.E-02 1.E-01 1.E+00 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce release to the environment MCA6 MCA7 MCA8 MCA9 Cerium aerosol release to the environment with a water spill

  • MCA6 and MCA7 had the higher releases release due the HVAC flow through the reactor building The auxiliary filter flow increased releases to the environment due to non-perfect capture by the absolute filters MCA6 and MCA7 were higher than the corresponding dry cases (MCA1 and MCA4) due to higher leakage from the reactor cell
  • MCA8 and MCA9 are essentially identical releases to the environment (explained on next slide)

MCA8 and MCA9 did not have the building HVAC flow Results of the sensitivity studies Cases with a water Spill Cases with a water spill Case Aerosol size Stack Fans Aux. Filters Water Spill MCA6 1 µm Yes No Yes MCA7 1 µm Yes Yes Yes MCA8 1 µm No Yes Yes MCA9 1 µm No No Yes

120 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 1.E-02 1.E-01 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce settling and filter behavior MCA8 condensing tank MCA8 airborne in the reactor cell MCA8 offgas MCA8 environment MCA9 condensing tank MCA9 airborne in the reactor cell MCA9 offgas MCA9 environment Environment MCA8 offgas Airborne in the reactor cell Condensing tank MCA9 offgas

(<10-11) 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 1.E-02 1.E-01 0

6 12 18 24 Release Fraction (-)

Time (hr)

Ce reactor building and filter behavior MCA7 condensing tank MCA7 offgas MCA7 environment MCA8 condensing tank MCA8 offgas MCA8 environment MCA8 offgas Condensing tank MCA7 offgas Environment Comparison of MCA7 and MCA8 shows the HVAC flow sweeps a portion of the small aerosols through the filters and out the stack

  • Most aerosols in the condensing tank
  • Stack flow capture and aerosol pass-through is more important than only the auxiliary filter flow Comparison of MCA8 and MCA9 shows the auxiliary filter has a negligible impact on the environmental release
  • Capture in the condensing tank, rapid settling in the reactor cell, and retention in the offgas system (filter pit and stack) overwhelms the importance of the auxiliary flow when the HVAC is not operating Results of the sensitivity studies Impact of HVAC flow with auxiliary filter flow Cases with a water spill Case Aerosol size Stack Fans Aux. Filters Water Spill MCA6 1 µm Yes No Yes MCA7 1 µm Yes Yes Yes MCA8 1 µm No Yes Yes MCA9 1 µm No No Yes Impact of auxiliary filter without HVAC flow MCA7 & MCA8 MCA8 & MCA9
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