ML25084A173
| ML25084A173 | |
| Person / Time | |
|---|---|
| Issue date: | 03/25/2025 |
| From: | Shawn Campbell NRC/RES/DSA/FSCB |
| To: | |
| References | |
| Download: ML25084A173 (1) | |
Text
SCALE & MELCOR non-LWR Source Term & Fuel Cycle Demonstration Project Heat Pipe Microreactor NRCs Volume 3 & 5 - Public Workshop March 26, 2025 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Office of Nuclear Material Safety and Safeguards Office of Nuclear Reactor Regulation 1
- NRC Strategy for non-LWR Readiness
- Microreactor Core & Fuel Cycle Overview
- Overview of the Simulated Accidents
- Demonstration of SCALE for Microreactor Core & Fuel Cycle Analysis
- Microreactor Modeling using MELCOR
- Summary & Closing Thoughts Outline 2
=
Background===
NRCs Strategy for Preparing for non-LWRs NRCs Readiness Strategy for Non-LWRs Phase 1 - Vision & Strategy Phase 2 - Implementation Action Plans IAPs are planning tools that describe:
Required work, resources, and sequencing of work to achieve readiness Strategy #2 - Computer Codes and Review Tools Identifies computer code & development activities Identifies key phenomena Assess available experimental data & needs IAP Strategy #2 Computer Codes and Tools Volume #1 Systems Analysis Volume #2 Fuel Performance Volume #3 Source Term, Consequence Volume #4 Licensing &
Dose Volume #5 Nuclear Fuel Cycle 4
Preparing for Non-LWR Licensing Near-Term Implementation Action Plan Strategy 1 Knowledge, Skills, Capacity Strategy 2 Analytical Tools Strategy 3 Flexible Review Process Strategy 4 Industry Codes
& Standards Strategy 5 Technology Inclusive Issues Strategy 6 Communication ML17165A069 ML24069A003 Briefed ACRS in April 2024 :
the NRC is in a good position to support technical reviews of advanced reactor design applications anticipated in the near future continued investments in [code] development and maintenance are necessary to ensure staff competency and effectiveness in technical reviews.
IAP Strategy 2 Volumes
6 Application to Licensing Reviews Developed Efficient Strategy and Plan Develop Reference Plant Models Expertise & Show Readiness Modify Reference Plant Model Support Hermes CP Review Current & Future Support 2018-2024 ML24069A003 HPR, June 2021 HTGR, July 2021 FHR, September 2021 MSR, September 2022 SFR, September 2022 MicroRx, March 2025
Pre-Application Activities for Advanced Reactors High Temperature Gas Reactors Light Water Reactors Molten Salt Reactors /
Molten Chloride Fast Reactors Sodium Cooled Fast Reactors Other Designs/ Not Specified Energy Northwest Deep Fission Abilene Christian University
Duke Energy - Belews Creek, NC General Atomics Electromagnetic Systems
- Last Energy Natura Resources TerraPower & GE - Hitachi Natrium Japan Atomic Energy Agency
- Radiant Industries, INC.
TerraPower, LLC Texas A&M University -
RELLIS Campus
- Terra Innovatum (SOLO)
TVA - Clinch River Nuclear Site Terrestrial Energy USA, INC.
- Westinghouse eVinci
- University of Illinois at Urbana-Champaign - NANO Nuclear Energy Inc.
Westinghouse AP300 X-Energy, LLC (XE-100)
X-Energy, LLC (XENITH)
- Indicates Microreactor Design
Demonstration Project:
Severe Accident and Source Term Analysis
Codes Supporting non-LWR Source Term and Fuel Cycle Licensing
- NRCs comprehensive neutronics package
- Nuclear data & cross-section processing
- Decay heat analyses
- Criticality safety
- Radiation shielding
- Radionuclide inventory & depletion generation
- Reactor core physics
- Sensitivity and uncertainty analyses
- NRCs comprehensive accident progression and source term code
- Characterizing and tracking accident progression,
- Performing transport and deposition of radionuclides throughout a facility,
- Performing non-radiological accident progression 9
10 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 The ultimate objective is to prepare the agency for effective and efficient licensing advanced reactor designs!
Project Objectives
11
- 1. Build SCALE core models and MELCOR full-plant models based off publicly available designs
- 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
12 5 MWth with a 5-year operating lifetime 1,134 heat pipes fueled with metallic U fuel (19.75 wt.% U-235)
Reactivity controlled via control drums
- 400 MWth reactor, graphite moderated
- Helium-cooled & TRISO-particle pebble-fueled at 10 wt.% U-235
- Fuel discharged at high burnup (90 GWd/MTIHM)
- 10 MWth reactor, graphite moderated at near atmospheric pressures
- Reactor fueled with liquid dissolved fuel in molten salt (34.5 wt. % U-235)
- 250 MWth pool-type reactor, utilizing metallic U / HT-9 fuel rods
- Reactor fueled with U-Pu-Zr fuel
- Liquid sodium coolant High-Temp. Gas Cooled Reactor PBMR-400 Sodium-Cooled Fast Reactor ABTR Molten Salt-Cooled Reactor UCB Mk1 PB-FHR Molten Salt-Fueled Reactor MSRE Heat Pipe Reactor INL Design A Public workshop videos, slides, reports at the NRC advanced reactor source term webpage
- 236 MWth reactor at atmospheric pressures
- Online refueling June 29, 2021 September 13, 2022 September 14, 2021 September 20, 2022 July 20, 2021 Previous Workshops
Demonstration Project:
Nuclear Fuel Cycle
Project Scope - Non-LWR Fuel Cycle Enrichment UF6 enrichment UF6 Transportation Fuel Fabrication Fresh Fuel Transportation Fuel Utilization (including on-site spent fuel storage)
- Not envisioned to change from current methods.
Uranium Mining & Milling
- Lack of information for non-LWR concepts Spent Fuel Off-site Storage &
Transportation
- Lack of information for non-LWR concepts Spent Fuel Final Disposal Stages in scope for Volume 5 Stages out of scope for Volume 5 14
Project Approach Build representative fuel cycle designs Identify key scenarios and accidents exercising key phenomena & models Build representative SCALE & MELCOR models and evaluate Code Assessment Representative Initial and Boundary Conditions Simulating Accidents around Key Phenomena Sensitivity Studies Identify &
Address Modeling Gaps 15
Representative Fuel Cycle Designs Completed 5 non-LWR fuel cycle designs for -
Heat Pipe Reactor (HPR)- INL Design A High Temperature Gas Reactor (HTGR) - Pebble Bed Modular Reactor (PBMR)-400 Fluoride-Salt Cooled High Temperature Reactor (FHR) - University of California, Berkeley (UCB) Mark 1 Molten Salt Reactor (MSR) - Molten Salt Reactor Experiment (MSRE)1 Sodium-Cooled Fast Reactor (SFR) - Advanced Burner Test Reactor (ABTR)
Identified potential processes & methods, for example:
What shipping package could transport HALEU-enriched UF6? What are the associated hazards?
How is spent SFR fuel moved? What are the hazards associated?
How is fissile salt manufactured for MSRs? What are the various kinds of fissile salt that may be used? What are the hazards?
Informed Initial and Boundary Conditions for the SCALE &
MELCOR Analyses ML24004A270 1The basis for the MSR fuel cycle design is based upon the MSRE, a small thermal spectrum reactor. Analyses showcased in this workshop utilized the MSR fuel cycle, but the facility and reactor input files were based on the geometry of the MSBR.
16
Previous Workshops 17 February 28, 2023 September 20, 2023 July 11, 2024
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 Overview of the TRISO-based Heat Pipe Microreactor Fuel Cycle
19 Microreactor Source Term and Fuel Cycle Leveraging previous source term and fuel cycle demonstration workshops:
TRISO fuel models - HTGR TRISO fabrication - HTGR Heat pipes models - HPR Heat Pipe Microreactor Fuel Cycle
20 Reference Heat Pipe Microreactor Design Refs.: Ortensi, J., et al., The Monolithic Heat Pipe Microreactor Reference Plant Model, INL/RPT-24-77914, Idaho National Laboratory, 2024 Reactor
- Based on a generic Heat Pipe Microreactor model (HPMR) [1]
- 7.5 MW thermal power
- 127 TRISO compact fuel assemblies
- 876 heat pipes
- 12 control drums with arcs of B4C for reactivity control
- Moderator is monolithic graphite block
- Power conversion is open-air Brayton cycle
21 Construction
- Metal pipe with wick along pipe inside surface
- Liquid coolant (sodium) fills area between wick and pipe inside surface Operation
- The core heats the liquid coolant which generates vapor
- The vapor flows to the other end of the heat pipe where it condenses, heating the secondary system fluid
- Coolant film return flow by capillary forces Heat pipe for reactor use
22 Reference Heat Pipe Microreactor Design TRISO Particle Heat Pipe Fuel assemblies
- 2 types of fuel assemblies 114 standard assemblies 13 control rod assemblies Outside of fuel element has hexagonal shape Kernel material is UCO within a graphite matrix/compact
- 876 heat pipes Sodium working fluid Wick and cladding of SS-316
23 Microreactor fuel cycle considerations New issues for heat pipe microreactors Higher enrichment impact on criticality during storage and transportation New chemicals and processes for fuel fabrication Sodium/potassium reactive with air and water Inert rooms and cells are common throughout the reactor system Sodium/potassium reactive with concrete Remote fuel handling systems introduces new challenges and hazard scenarios Flexibility in scenario investigation Conservative or best-estimate boundary conditions Uncertainty assessment - identification of safety margins
24 Transportation of microreactor Fresh core staging/
installation Power production On-site storage TRISO-based Heat Pipe Microreactor fuel cycle
- Enrichment
- Gas centrifuges Transportation of UF6 (or other forms)
Fabrication of HPR fuel compact, fuel block, and core E1 T1 F1 F2 T2 U1 U2 U4
- TRISO fabrication
- Sol-gel process
25 E1: Enrichment US facilities
Uranium enrichment - both gas centrifuge o
Louisiana Energy Services in Eunice, NM Only active commercial process for LEU enrichment in the United States Announced in 2019 plans for on site for enrichment, deconversion, & fuel fabrication o
Centrus Energy Corp in Piketon, OH Announced as first HALEU facility on June 23, 2021 20% enrichment - $170M DOE project to make 600 kg by June 2022
26 Orano DN30-X package for up to 20% enrichment:
License was approved by the NRC in March 2023
30B-X cylinder similar to 30B cylinder, but with criticality control features (absorber rods)
30B cylinder:
30 inches diameter, capacity of 2277 kg UF6
Licensed up to 5% U-235 enrichment
DN30-10 and DN30-20
DN30-10: capacity 1460 kg UF6 with enrichment up to 10%
DN30-20: capacity 1271 kg UF6 with enrichment up to 20%
Fuel vendors may desire to ship other fuel forms in other packages T1: Transportation of UF6 Orano 30B cylinder with B4C rods DAHER DN30: 30B cylinder with PSP Due to differences in various HTGR/FHR designs, a range of enrichments between 5-20% will be considered.
27 F1: Fabrication of TRISO particles Fuel kernel:
U.S. TRISO production based on internal sol-gel process
Starting sol is a uranyl nitrate solution
Sol is dripped through a nozzle into a heated organic diluent (silicone oil)
Heat causes HMTA (Hexamethylenetetramine) to chemically decompose and induces a gelation reaction which eventually forms the fuel kernel
Reference:
Peter Pappano, TRISO-X Fuel Fabrication Facility Overview, Introductory Meeting with the NRC, ML18254A086 (2018). Obtained from https://www.nrc.gov/docs/ML1825/ML18254A086.pdf on October 20, 2021.
28 F2: Fabrication of Fuel Compact - Boundary conditions
Graphite powder is dried, pulverized and then is used for overcoating the TRISO kernels at controlled temperatures
Compaction of fuel zone
Final step includes carbonization and heat treatment before compact is released for inspection
Reference:
Paul A. Demkowicz, TRISO Fuel: Design, Manufacturing, and Performance, INL/MIS-19-52869-Revision-0 (2019).
29
Each fuel compact will be assembled at the fabrication facility
Heat pipes can be fabricated in a separate factory with material testing and inspection of the stainless steel pipe, metal pipe and wick
Assembly of heat pipe components and filling of the heat pipe with sodium can be done in a temperature-and pressure-controlled environment F2: Assembly of Fuel Compacts and Heat Pipe -
Boundary Conditions Fuel compact fab Heat pipe fab Assembly
30 T2: Transportation of Fresh Core
Versa-Pac, which is used to transport fuel pebbles, could be used for fuel compacts transportation to the fuel assembly fabrication facility
On the other hand, microreactor and its fuel can be transported separately or together in a cargo container by highway, rail, or ship.
Setting boundary:
Microreactor and its fuel are transported together with vehicle in a CONEX shipping container
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 Selected Scenarios
32 Frontend Fuel Cycle Criticality event with fresh core Power Production Unprotected reactivity insertion Loss of heat removal Backend Fuel Cycle Criticality event with spent core Shielding analysis with spent core in vehicle accident Vehicle accident with a breach in containment Scenarios
33 Description of SCALE Scenarios Scenario 1 Criticality event during fresh fuel transportation Base Case:
- The fresh reactor core is immersed in a body of water Base Case
- The irradiated reactor core is immersed in a body of water Scenario 2 Criticality event during irradiated fuel transportation Scenario 3 Shielding analysis during spent fuel transportation Water Base Case:
- Dose assessment in area surrounding the reactor
34 Description of MELCOR Scenarios Scenario 1 Unprotected Reactivity Initiated Accident Base Case:
- Inadvertent rotation of control drums near end of cycle causing insertion of reactivity Base Case
- Trip in turbine/compressor causing loss of heat removal in heat exchanger Scenario 2 Unprotected Loss of Heat Sink Base Case
- Collision during spent core transport resulting in package failure and fire Scenario 3 Transport Accident
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 Demonstration of SCALE Analysis for Scenarios in TRISO-fueled Heat Pipe Microreactor Fuel Cycle D. Hartanto, G. Radulescu, F. Bostelmann, W. A. Wieselquist
36 OBJECTIVE AND APPLICATIONS
- Initiating event: Truck with fresh core drives into river
- Analysis: Perform criticality calculations with and without water ingress and control rod displacement Scenario 1: Criticality event during fresh fuel transportation
- Initiating event: Vehicle accident with spent core drives into river
- Analysis: Perform criticality calculations with and without water ingress and control rod displacement Scenario 2: Criticality event during spent fuel transportation
- Initiating event: Vehicle accident resulting in loss of package shielding
- Analysis: Dose analysis with and without shielding Scenario 3: Shielding analysis during spent fuel transportation Objective: Demonstrate the use of SCALE for simulating accident scenarios in all stages of the nuclear fuel cycle for TRISO-fueled heat-pipe microreactors (HPMRs)
HPMR Model
37 The HPMR SCALE model was developed based on the monolithic generic HPMR model developed by Idaho National Laboratory [1] with several adjustments to achieve a substantial core lifetime, such as using a BeO reflector REFERENCE HPMR MODEL
Reference:
[1] Ortensi, J., et al., The Monolithic Heat Pipe Microreactor Reference Plant Model, INL/RPT-24-77914, Idaho National Laboratory, 2024 Side view - SCALE HPMR Model Parameter Value Core power 7.5 MWth Core lifetime 3 EFPYs Reflector BeO Monolith Graphite Number of heat pipes 876 Fuel assembly (FA) types 2
Number of standard FAs 114 Number of control rod FAs 13 FA Pitch 17.368 cm Pin Pitch 2.782 cm Number of control drums 12 160 cm 20 cm 20 cm SS-316 Graphite Heat pipe Helium 293.6 cm Reflector EFPY: Effective Full Power Year
38 The tristructural-isotropic (TRISO) particles and the compact designs were based on the AGR-2 design [2]
REFERENCE HPMR MODEL
Reference:
[2] A. Sowder and C. Marciulescu, Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO)-Coated Particle Fuel Performance, Electric Power Research Institute, Palo Alto, CA, Topical Report EPRI-AR-1(NP)-A 3002019978, 2020.
Top view - SCALE HPMR Model TRISO particles in a compact Parameter Value Kernel material UCO Kernel enrichment 19.75 wt% 235U Kernel density 10.5 g/cm3 Kernel radius 0.02125 cm Coating layer material C/iPyC/SiC/oPyC Coating layer thickness 100/40/35/40 µm TRISOs packing fraction 40%
Compact fuel zone radius 0.875 cm Compact fuel zone height 2.45 cm Compact non-fuel zone radius 0.90 cm Compact non-fuel zone height 2.50 cm Compact axial stack per pin 64 Standard FA (24 pins)
Control rod FA (18 pins)
IPyC: inner pyrolytic carbon OPyC: outer pyrolytic carbon
39 APPLIED SCALE 6.3.2 SEQUENCES Reactor physics burnup calculations with TRITON (TRITON-Shift)
Shielding & radiation dose calculations with MAVRIC-Shift
- Monte Carlo photon and neutron transport code (Shift) with automated variance reduction for shielding analyses
- Requires radiation source terms
- Output:
- Spatial flux and dose rate distributions Criticality calculation with CSAS (CSAS-Shift)
- Monte Carlo neutron transport code (KENO or Shift) for criticality analysis
- Output:
- Multiplication factor
- Spatial flux and fission density distributions
References:
[3] Wieselquist, W. A., Lefebvre, R. A., Eds., SCALE 6.3.1 User Manual, ORNL/TM-SCALE-6.3.1, Oak Ridge National Laboratory, 2023.
[4] Chadwick, M. B., et al. ENDF/B-VII. 1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data. Nuclear data sheets, 112(12), 2887-2996, 2011.
- Monte Carlo neutron transport code (KENO or Shift) coupled with ORIGEN depletion solver
- Output:
- Effective multiplication factor
- Spatial flux and power distributions
- Nuclide inventory ENDF/B-VII.1 [4] nuclear data library used for all calculations
40 Monte Carlo codes in SCALE (CSAS and TRITON) support the neutron transport calculation using either continuous energy (CE) and multigroup (MG) cross sections for double-heterogeneity problems MG calculations reduce computing time while maintaining accuracy, useful especially for depletion calculation DEPLETION CALCULATIONS Illustrations of the explicit TRISO particles model (left) and its equivalent double-heterogenous mixture model (right) in SCALE Case Library Keff Difference (pcm)
Speed-up factor Absorber in control drum facing the core CE 0.99897 +/- 0.00031 MG-252g 0.99972 +/- 0.00022 75.6 +/- 38.0 6.65 Absorber in control drum facing outwards CE 1.05522 +/- 0.00026 MG-252g 1.05594 +/- 0.00017 72.3 +/- 31.1 7.69 Comparison between CE and MG of keff of HPMR 3D core
41
- Depletion cases performed with SCALE/TRITON:
All control absorbers drum rotated outwards during depletion Time-dependent control drums position
Driven via a Python script
Polynomial fitting and brentq root finding method to determine the angle of control drums
Initial keff at beginning of depletion time step was set to 1.004 to remain critical in each depletion time step (30 days)
Total of 112 depleted regions (7 radial and 16 axial regions)
Leveraged new flexibility in SCALE/TRITON (SCALE 7.0 beta): avoid traditional heavy metal mass normalization and use total reactor power instead of specific power DEPLETION CALCULATIONS Evolution of keff as function of days/burnup w/
and w/o time-dependent control drums position EFPD: Effective Full Power Day GWd/MTIHM: Gigawatt-day / Metric Ton of Initial Heavy Metal Outward (0)
Inward(180)
42 COMPOSITION DISTRIBUTION IN SPENT FUEL Less fission products were produced in HPMR (1.48%) compared to the PWR (4.55%) due to lower discharged burnup HPMR achieved burnup of 15.88 GWd/MTIHM after 3 years of operation and PWR discharged fuel was taken from burnup of 50 GWd/MTIHM The HPMR depletion cases with and without time-dependent control drum positions had similar composition distributions Consideration of time-dependent control drum position may slightly harden the spectrum At discharged, 239Pu mass is about 3%
increased AC: actinides, CD: control drum, FP: fission products, LT: light elements, PWR: pressurized water reactor Comparison of composition distribution in spent fuel PWR fuel with initial enrichment of 4.2 wt% achieving burnup of 50 GWd/MTIHM
43 Decay heat at shutdown is about 6% operating power in both the HPMR and PWR Actinides, such as 239U and 239Np, are not found among top 5 decay heat contributors of HPMR due to its low burnup DECAY HEAT OF SPENT FUEL Top 5 contributors at shutdown for HPMR Top 5 contributors at shutdown for PWR 134I (1.94%)
239U (2.73%)
138Cs (1.87%)
239Np (2.39%)
92Rb (1.77%)
134I (1.89%)
144La (1.72%)
138Cs (1.76%)
91Rb (1.66%)
104Tc (1.64%)
Top 5 contributors after 5-year cooling for HPMR Top 5 contributors after 5-year cooling for PWR 90Y (36.09%)
134Cs (18.78%)
137mBa (26.05%)
90Y (18.54%)
144Pr (8.17%)
137mBa (18.38%)
90Sr (7.57%)
244Cm (8.37%)
137Cs (7.46%)
106Rh (7.80%)
Comparison of decay heat at discharged between HPMR and 50 GWd/MTIHM PWR
44 Reactivity coefficients and kinetics parameters were calculated by SCALE/CSAS-Shift and provided to MELCOR Temperature reactivity coefficients are negative across the entire temperature range at both BOL and EOL REACTOR PHYSICS PARAMETERS Temperature feedback reactivity coefficients of fuel and moderator at BOL and EOL Parameter BOL EOL Fuel temp. coef.
(pcm/K)
-4.27 +/- 0.07
-4.37 +/- 0.07 Mod. temp.
coef. (pcm/K)
-1.42 +/- 0.13
-1.13 +/- 0.15 Prompt neutron gen. time (s) 1.875E-04 1.920E-04 Beta-eff 6.594E-03 +/-
9.882E-05 6.078E-03 +/-
1.042E-04 BOL: Beginning of Life, EOL: End of Life
45 Scenario 1: Truck with fresh core drives into river Criticality event during fresh fuel transportation
46
- The reactor core is immersed in the center of a body of water with a thickness of 100 cm
- The optimal surrounding water thickness is about 10 cm in a full immersion scenario CRITICALITY MODEL FOR FRESH CORE Criticality model of HPMR core immersed in water Control rod (B4C, nat. enrich)
Water Top view Side view Control drum absorber (B4C, nat. enrich)
Water Multiplication factor as a function of surrounding water thickness for full immersion case with control drums and control rods inserted
47 The high excess reactivity at fresh condition requires the negative reactivity from both control rods and control drums to bring the core to subcritical (keff < 0.95) during water immersion The current core configuration requires a higher worth of control mechanism to prevent criticality due to water immersion CRITICALITY ANALYSIS FOR FRESH CORE Case Water immersion Control rods position Direction of absorber of control drum A-1
- Partial Out Outward A-2 Inward A-3 In Outward A-4 Inward B-1 Full Out Outward B-2 Inward B-3 In Outward B-4 Inward
- Water does not enter the active core region such as the fuel compact gap and control rod tube Simulated criticality cases Multiplication factor of simulated criticality cases for fresh fuel
48 Replacing natural boron in the control rods by enriched boron provides approximately 2000 pcm of negative reactivity, but it still does not yield a keff < 0.95 for the fully immersed case CRITICALITY ANALYSIS FOR FRESH CORE Multiplication factor for full immersion with control drums and enriched control rods inserted Comparison of neutron spectra in kernel at normal condition and when immersed in water
49 Scenario 2: Vehicle accident with spent core drives into river Criticality event during irradiated fuel transportation
50 The same model as the criticality model for fresh core was used Fuel compositions were taken immediately after discharge, and after 1 year and 5 years of cooling In addition to C and O, only key nuclides relevant to burnup credit were included in the fuel composition, following the recommendations of NUREG-2216 [5]
CRITICALITY MODEL FOR SPENT CORE Criticality model of HPMR core immersed in water
Reference:
[5] J. Borowski et al., Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, NUREG-2216, 2020 Type Set of Nuclides Actinide-only burnup credit 234U, 235U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am Additional nuclides for actinide-plus-fission product burnup credit 95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 133Cs, 143Nd, 145Nd, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 151Eu, 153Eu, 155Gd, 236U, 237Np, 243Am Water
51 Both shutdown rods and control drums are still required to guarantee the critical safety limit of the irradiated core during water immersion CRITICALITY ANALYSIS FOR SPENT CORE Case Water immersion Control rods position Direction of absorber of control drum A-1
- Partial Out Outward A-2 Inward A-3 In Outward A-4 Inward B-1 Full Out Outward B-2 Inward B-3 In Outward B-4 Inward
- Water does not enter the active core region such as the fuel compact gap and control rod tube Simulated criticality cases Multiplication factor of simulated criticality cases for irradiated fuel
52 Scenario 3: Vehicle accident resulting in loss of package shielding Shielding analysis during spent fuel transportation
53 OPERATIONAL RADIATION DOSE RATE Space-and energy-dependent prompt neutron fission source from a TRITON-Shift calculation Tally mesh: 40 x 40 x 50 voxels Number of neutron energy groups: 200 Total neutron source strength (normalized to a power of 7.5 MWth): 5.774E+17 n/s Prompt gamma source generated by the MAVRIC-Shift shielding code based on prompt fission neutrons Fission product decay and activation sources not included in the demonstration shielding calculation (the source component is lower than the prompt neutron and gamma sources)
Fission source spatial distribution TRITON-Shift or CSAS6-Shift
- Save fission source into a HDF5 output file MAVRIC-Shift
- Read fission source file from HDF5 output file
- Dose calculation Total fission neutrons per voxel volume
54
- Assumed outer core shielding model: 30 cm stainless steel shield
- Assumed reactor enclosure: 1.5 m concrete walls and shield door with layered materials Concrete wall, 150 cm thick Concrete wall, 150 cm thick Concrete floor, 150 cm thick 300 cm 350 cm Concrete wall, 150 cm thick 200 cm Steel shield, 30 cm thick OPERATIONAL RADIATION DOSE RATE SCALE model for radiation shielding calculation Front view Side view Concrete floor, 150 cm thick Boron steel, 5 cm thick Pb, 25 cm thick
- B4C, 25 cm thick
55 OPERATIONAL RADIATION DOSE RATE Dose (rem/h)
- The dose rate is <0.5 rem/h at any of the outer building regions Max dose:
0.5 rem/h Max dose:
0.3 rem/h
56 DOSE RATE DURING TRANSPORTATION Assumed simple package for transportation: 27 cm carbon steel and 13 cm resin [6]
Sources of radiation from 1 year and 5 years of cooling:
- Fission product and actinide decay
- Stainless steel activation assuming a 500 ppm cobalt impurity concentration in the heat pipe envelope In the event of a hypothetical accident, the part of the package within the red-dotted region is assumed to open (beyond-design-basis event) 10 CFR 71 requirements:
Normal condition: 10 mrem/h at any point 2 m from the outer lateral surfaces of the vehicle Hypothetical accident condition: 1 rem/h at 1 m from the external surface of the package Carbon steel, 27 cm thick
- Resin, 13 cm thick SCALE model for transportation package dose rate calculation
Reference:
[6] I.C. Gauld and J.C. Ryman, Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel, NUREG/CR-6700, 2001 Soil
57 DOSE RATE DURING TRANSPORTATION Cooling Time (years)
Dose at 2 m
/ normal condition (mrem/h)
Dose at 1 m /
hypothetical accident (rem/h) 1 0.05 1.21 5
0.02 0.67 Dose (rem/h)
Radiation dose map at normal condition Radiation dose map during hypothetical accident 10 CFR 71 requirements:
Normal condition: 10 mrem/h at any point 2 m from the outer lateral surfaces of the vehicle Hypothetical accident condition: 1 rem/h at 1 m from the external surface of the package
58 Requirement from 10 CFR 73.37 for physical protection of irradiated fuel in transit: 1 Gy/h (100 rad/h) at 1 m distance from any accessible surfaces without intervening shielding Sources of radiation:
- Fission product and actinide decay
- Stainless steel activation assuming a 500 ppm cobalt impurity concentration in the heat pipe envelope DOSE CHARACTERISTICS Thin air region for tally at 1 m from reactor surface Unshielded HPMR SCALE model
59 The HPMR has dose characteristics above 100 rad/h for about 7.7 years (2,800 days) after shutdown The PWR fuel assembly has dose characteristics above 100 rad/h longer
(>100 years) because of higher burnup of 50 GWd/MTIHM although the fuel volume is similar between PWR fuel assembly and HPMR core Single PWR fuel volume is about 92% of the total fuel volume in the HPMR core unit DOSE CHARACTERISTICS Maximum dose rate at 1 m distance of HPMR core unit and 50 GWd/MTIHM PWR fuel assembly
60 Summary
61
- SCALE capabilities to simulate different scenarios in the different HPMR fuel cycle stages were demonstrated through the calculation of Fuel inventory Decay heat Shielding and radiation dose Criticality
- Key observations:
Considering the time-dependent positions of control drums can slightly harden the spectrum, leading to a 3% increase in 239Pu buildup The significant excess reactivity in the fresh condition requires a large negative reactivity contribution to meet the critical safety limit during a water immersion accident Leveraging detailed information about the reactor enclosure and transportation package will enhance the accuracy of radiation dose calculations With a discharged burnup of 15.88 GWd/MTIHM, the HPMR core exhibits dose rate above 100 rad/h for approximately 7.7 years (2,800 days) after shutdown
SUMMARY
62
- Related recent development efforts in SCALE to enable explicit modeling of a higher packing fraction for TRISO-based fuels (up to 55%)
- Planned future enhancements:
Parallelization of the SCALE cross-section processing module (XSPROC) to dramatically reduce total MG calculation runtime for complex TRISO models Critical control element position search during depletion Calculation of adjoint-weighted kinetics parameters Reduction of the memory footprint of continuous-energy depletion calculations for models using random TRISO particle distributions FUTURE DEVELOPMENT
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 Energy's National Nuclear Security Administration under contract DE-NA0003525.
MELCOR TRISO Fueled Heat Pipe Reactor Model SAND2025-02818PE
64 OBJECTIVES
- TRISO fueled, horizontal Heat Pipe Reactor (HPR): Model development of a generic TRISO fueled horizontal Heat Pipe Micro-Reactor (HPMR) based on pre-conceptual, non-proprietary designs from the literature.
Develop analysis framework and insight on HPR performance and safety Demonstrate modeling capabilities with MELCOR
- Current modeling capabilities into generalized framework: Generalized representation for RN and COR package to lift current capabilities into modernized framework, enhanced flexibility/degrees of freedom Models developed and utilized
65 APPLICATIONS Plant Model
- Steady state conditions: Expected operational performance with generalized modeling capabilities (fission product migration, heat pipe operational envelope)
- Design basis events/accidents & sensitivities:
Passive safety response from design basis and beyond design basis accidents, impact on response due to varying severity of accident
- Margin assessment evaluations: Considering variable reactor system performance to exercise modeling capabilities Operational occurrences and transient events Transport Model
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 Energy's National Nuclear Security Administration under contract DE-NA0003525.
MELCOR Development
67
- MELCOR packages initially centered on LWR models Rigid structure and connection of core components, recycled for non-LWR designs Cumbersome to develop new physics models
/ component relationships for complex reactor designs
- Generalizing solutions in the COR & RN packages Eliminate recycling relationships to enhance flexibility for advanced non-LWR reactor designs Allow more degrees of freedom for modeling with COR and RN Generalized COR and RN packages Predecessor to modernized RN, Generalized Radionuclide Transport and Release model (SFR Vol. Workshop)
68 MELCOR accommodates heat pipe models of different fidelity through a common interface and a specified wall and working fluid region nodalization.
- Model 1: working fluid region modeled as high thermal conductivity material.
- Model 2: Thermodynamic equilibrium of working fluid. Pressure, temperature, and liquid/vapor fraction evolve in time. Sonic, capillary and boiling limits enforced.
COR Heat Pipe Modeling fission gas plenum region BeO reflector region lower head lower support plate outer baffle plate inner baffle plate alumina reflector core barrel neutron shield Ring 1 - Control Rod Rings 2-15 are the active core (each ring = pitch of 1 fuel element)
Ring 1 is the control rod guide Reflector and neutron shield Evaporator (fuel elements)
Levels 3-12 Condenser (secondary heat exchanger)
Level 14 Lower reflector Levels 1-2 Level 13 INL design type A HPR COR model MELCOR heat pipe model
69 Generalized COR Heat Pipe Modeling (x7) heat pipe (x24) compact heat pipe center heat pipe clad compact TRISO (x49765)
OPyC SiC IPyC buffer kernel assembly gap fuel assembly TRISO heat pipe Ring 2 Ring 1 fuel block Generalized COR package allows user specified components Allows for any number of components Enhanced flexibility for advanced, non-LWR reactor designs Simple geometric specification, property definitions, and physics models Framework for generalized components simplifies user input Reduces input requirements / less error prone for users Allows for rapid modeling iteration Ring 2 Ring 1 Heat Exchanger Reflector Reactor Evaporator Adiabatic Condenser heat Ring 1 control volume Ring 2 control volume heat
70
- Heat pipes have operational limits based on incident power and temperature Sonic limit:
Maximum velocity of vapor from evaporator to condenser Capillary limit:
Liquid flow rate at maximum capillary pressure difference Entrainment limit:
Entrainment of the working fluid diminishes the return flow in a similar manner as the capillary limit Boiling limit:
Onset of diminished heat transfer capacity due to nucleation in the heat pipe wick
- Limit curves are specified through input Curves developed from LANL code, HTPipe Heat Pipe Limits INL Design Type A HPR HPMR Heat pipe operating conditions Heat pipe operating conditions
71 Fission Product Diffusion Compact-Assembly Gap Compact OPyC SiC IPyC Buffer Kernel Radionuclide path for HPMR is assumed to travel from TRISO kernel, layers, graphite compact, and into compact-assembly gap For failed TRISO, fission products migrate from kernel to graphite compact Diffusivity coefficients specified through input to predict fission product release from fuel 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 throughout system (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 Reduce heat capacities for structures to accelerate approach to steady state thermal conditions.
= 1
+
= 0
0.00001 0.0001 0.001 0.01 0.1 1
1200 1400 1600 1800 2000 2200 2400 2600 Failure Fraction (-)
Temperature (°C)
TRISO Failure Fraction vs Temperature Specified failure versus temperature
Why is it important?
- Integration allows for radionuclide transmutation/decay calculations alongside MELCOR transport calculations
- Radionuclide decay and chemical-physical transformation impacts fission product transport
- Better estimates of:
- Fission product retention barrier efficacy (fuel, primary system, building/containment)
- Decay heat distribution
- Radiological hazards to personnel and equipment
- Ultimate source term to environment
- Reactor kinetics and neutronics feedback
- Isotopic transmutation and decay has implications for reactor kinetics (transients without scram)
- Important components of neutronic feedback depend on build-and-burn dynamics of key isotopes ORIGEN/MELCOR Integration Vol. 5: MSR Fuel Cycle Workshop
73 Enhanced RN package capabilities with MELCOR/ORIGEN integration Species transport formulated generally in terms of mass fluxes Interspatial Transport: Transport of species between volumes hosting species Aerosol Dynamics: Transformation of species within hosting volumes associated with vapor and aerosol dynamics Nuclear Processes: Transformation of species within hosting volumes due to nuclear transmutation and decay Chemical Processes: Transformation of species within hosting volumes due to chemical reactions RN Package Generalized Dynamics Transmutation/Decay Chemical Reactions Spatial Transport New Species State Initial Species State Aerosol Dynamics Other Transformations Initiation Transformation (Integrate over time step)
Solution
- ORIGEN calculates the transmutation of isotopes
- MELCOR calculates how isotope mass moves around a facility due to various mass transport processes
- MORTY used to establish connections between ORIGEN and MELCOR
- Recently enhanced to provide Xenon feedback and respond to power or flux variations Approach for ORIGEN/MELCOR Integration Transmutation &
decay Updated inventory Xenon reactivity feedback Power or neutron flux Point kinetics Updated power ORIGEN MORTY MELCOR Temperature dependent flux profile Isotopic inventory
75
- Within MELCOR, radionuclide classes consist of elements grouped together based on commonality (chemical or physical transport)
Krypton, Cesium, Strontium, and Silver classes for radionuclide diffusion in TRISO With integration, transmutation contributes to mass transfer between classes
- ORIGEN informs MELCOR on Xenon worth based on changes in isotopic concentration MELCOR passes power variation to ORIGEN ORIGEN calculates new Xenon concentrations from transmutation and destruction ORIGEN passes Xenon worth back to MELCOR for point kinetics calculations
- User input polynomial coefficients to fit mass to worth (reactor specific)
MELCOR with ORIGEN Integration Class Name Chemical Group Element Members XE Noble Gas He, Ne, Ar, Kr, Xe, Rn, H, N CS Alkali Metals Li, Na, K, Rb, Cs, Fr, Cu BA Alkaline Earths Be, Mg, Ca, Sr, Ba, Ra, Es, Fm AG Less Volatile Ga, Ge, In, Sn, Ag
Xenon-135 concentration to worth
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 Energy's National Nuclear Security Administration under contract DE-NA0003525.
MELCOR Modeling Description
77
- 2D nodalization 7 rings and 8 axial levels for active core Each ring being a set number of full fuel assemblies
- Control volume for each ring, spanning from core to heat exchanger, represent heat pipes
- One control volume to represent all interstitial gas volumes in core Assembly gaps, pellet-assembly gaps, control rod voids
- 1D cylindrical heat transfer between heat pipe fluid region and heat exchanger MELCOR Nodalization ring 1 ring 2 ring 7
Fuel assembly Control rod assembly
78 Iteration on core design to achieve cycle length and power output objectives 40% TRISO packing fraction, BeO reflectors, 7.5 [MW] thermal power Power distribution from SCALE calculations SCALE calculations in support of transient analysis with MELCOR Reactivity temperature feedback & kinetics parameters Isotopic inventory at End Of Cycle (EOC)
Xenon concentration vs worth coefficients for Xenon feedback 252 energy group flux distribution for isotope depletion SCALE Inputs Power distribution Reactivity as a function of temperature Kinetics Parameters Value (EOC)
1.920E-04
6.078E-03 1l 1 1.271E-02 l3.751E-02 2l 2 3.167E-02 l2.139E-01 3l 3 1.162E-01 l1.880E-01 4l 4 3.123E-01 l4.016E-01 5l 5 1.395E+00 l1.276E-01 6l 6 3.839E+00 l2.691E-02
=
+
=
EOC isotopes and decay heat 252 energy group flux distribution Xenon-135 concentration to worth
79 Model for Operational Phase
- Open air Brayton cycle Built in MELCOR mechanical models used for turbine/compressor modeling 10x pressure ratio and 200 [K]
temperature increase by compressor
- Reactor resides above grade in reactor core room Vents from room to environment to prevent core room heat up Room encased in concrete with steel door entrance
- Core room isolated from power conversion unit Reactor room vents
80 Reactor vessel encased in a steel and epoxy layer during transport Assuming LWR spent fuel transport package specifications 3-year nominal operation with decay heat cool down time prior to transport Reactor core not depressurized for transport Model for Transport Phase Pressure boundary Reactor core
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 Energy's National Nuclear Security Administration under contract DE-NA0003525.
MELCOR Steady State and Transient Analysis
82
- Nominal power selected to reach target cycle length Relatively low power density core
- Steady state model used in accelerated diffusion calculation Predicts fission product distribution after 3-year nominal operation Steady State Results Parameter Value Thermal Power [MW]
7.5 H.X. Temp. [K]
845 Max. Fuel [K]
1436 Max. Assembly [K]
1400 Avg. Fuel [K]
1353 Avg. Assembly [K]
1325 Fuel kernel and heat pipe coolant temperatures
- TRISO/Compact diffusivities from IAEA TECDOC-978 Classes released into core interstitials Low SiC layer diffusivity Low kernel layer diffusivity
83 Description of Transient Scenario 1:
Unprotected Reactivity Initiated Event Inadvertent rotation of control drums near end of cycle causing insertion of reactivity No intervention or protection systems triggered Trip in turbine/compressor causing loss of heat removal in heat exchanger No intervention or protection systems triggered Scenario 2:
Unprotected Loss of Heat Sink Event Collision during spent core transport Rupture in package and pressure vessel Package subjected to fire Scenario 3:
Transport Accident Reactor room vents
84 Base Case:
0.25 [$] reactivity insertion from 0 to 100 seconds near EOC
- Strong passive safety features Large negative temperature feedback Large heat pipe operational boundary Low energy density, high thermal inertia Scenario 1:
Unprotected Reactivity Initiated Event Power and average fuel/assembly temperatures
85 Sensitivities case 1&2:
Amount of reactivity inserted Variable amount of reactivity inserted over 100 seconds Scenario 1:
Unprotected Reactivity Initiated Event Excess reactivity vs cycle length Sensitivities case 3&4:
Duration of reactivity inserted 0.25 [$] of reactivity inserted over variable time Inserted reactivity vs time 1.0 [$], 980 [EFPD]
~90% of cycle length 0.50 [$], 1051 [EFPD]
~95% of cycle length 0.25 [$], 1086 [EFPD]
~99% of cycle length
86 Scenario 1:
Unprotected Reactivity Initiated Event Power and average fuel temperatures Power and average fuel temperatures Sensitivities case 1&2:
Amount of reactivity inserted Sensitivities case 3&4:
Duration of reactivity inserted Heat pipe and fuel integrity not expected to be compromised for any of these cases
- Larger reactivity insertions lead to higher powers and temperatures
- Shorter reactivity insertions lead to slightly higher powers, but similar temperatures
87 Scenario 1:
Unprotected Reactivity Initiated Event Heat pipe limit curves 10x reduced boiling limit Sensitivity cases 5-10 Heat pipe w/ reduced boiling limit Base case (0.25 [$] over 100 [s]) but with N number of heat pipes with reduced boiling limit in center assembly Exceeding boiling limit causes reduced heat transfer Heat transfer limited to 25%
Full heat transfer restored if heat pipe drops below boiling limit Heat Pipes Cross section of center assemblies Heat pipe operating conditions Center assemblies
88 Scenario 1:
Unprotected Reactivity Initiated Event Center heat pipe temperature Center fuel temperature Heat pipe limit curves (zoomed in)
- With reduced boiling limit, heat pipes would surpass limit leading to reduced heat transfer
- More heat pipes with reduced boiling limits means higher and more rapid increase in temperatures Increasing number of heat pipes w/ reduced boiling limit Increasing number of heat pipes w/ reduced boiling limit Return from boiling limit Heat pipes during nominal operation Higher temperatures when more HPs surpass boiling limit Critical number of heat pipes with reduced boiling limits lead to significant loss of heat removal
89 Scenario 2:
Unprotected Loss of Heat Sink Event Base Case:
Trip in turbine/compressor resulting in loss of forced flow in heat exchanger No intervention or protection systems activated Power and average fuel/assembly temperatures Coupling of thermal hydraulic and neutronic behavior identified reactivity excursion due to reactor cool-down Loss of forced coolant flow increases temperatures Gradually reduces core component temperatures due to passive heat removal Around 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> into transient, power excursion due to component temperatures cooling down Reactor room vents
90 Scenario 2:
Unprotected Loss of Heat Sink Event Sensitivities case 1&2 Variation in flow rate for reactor room vents
- Reducing flow rate of reactor room vents causes reactor room to heat up, reduces heat removal from core
- Less heat removal leads to higher core component temperatures Power and average fuel temperatures Degraded reactor heat removal inhibits occurrence of reactivity excursion Reactor room vents
91 Scenario 2:
Unprotected Loss of Heat Sink Event Sensitivities case 1&2 Variation in flow rate for reactor room vents HPMR design has a very low power density and high thermal capacity of core components (BeO reflector &
graphite assemblies)
Temperatures rise very slowly Fission product release into interstitials dominated by steady state operation CS class in core interstitials AG class in core interstitials BA class in core interstitials Compact-assembly interstitial Assembly interstitial Control rod interstitial Low energy density of system and large thermal inertia causes radiological release to be dominated by operational leakage out of TRISO fuel particles even under accident conditions
92 Unlike traditional reactor designs, this reactor concept has a very short iodine pit Scenario 2:
Unprotected Loss of Heat Sink Event with MELCOR/ORIGEN for Xenon Feedback
= +
()
Change in Xe-135 for various I-135 to Xe-135 ratios Flux and Xe-135 abs. cross section I-135 Xe-135 Cs-135 Created from
- fission, destroyed by decay Decay Decay Created from
- decay, destroyed by decay and neutron absorption Created from
- decay, destroyed by decay Decay
+
93
- Xe-135 concentration quickly begins to drop below equilibrium concentration Low equilibrium I-135/Xe-135 ratio and no more production of I-135 due to power decrease Scenario 2:
Unprotected Loss of Heat Sink Event with MELCOR/ORIGEN for Xenon Feedback Xe-135 change from initial concentration What is causing the temporal dynamics?
Total Power
- Xe-135 concentration dropping below equilibrium will lead to a positive reactivity insertion Xe-135 is a strong neutron absorber/poison Less Xe-135, more neutrons for fission Modeling Xe-135 feedback identifies earlier and larger power excursion
94
- Low power oscillations caused by component temperature feedback
- As component temperatures cool down, they contribute less negative excess reactivity Scenario 2:
Unprotected Loss of Heat Sink Event with MELCOR/ORIGEN for Xenon Feedback Power and Temperatures Reactivity Without Xe feedback With Xe feedback Multi-physics modeling critical to capture realistic reactor response for unprotected loss of heat sink event
95 Base Case
- Spent core transported after 3-year operation, 1-year cool down
- Package is severely damaged during transport, large flow area between heat exchanger region and environment
- Reactor pressure boundary is damaged, leaks around heat pipe, control rod, and control drum penetrations
- Subjected to external fire,
~1100[K] for 60+ minutes Scenario 3:
Transport Accident Sensitivities case 1&2:
Variation in reactor cool down time
- Time allotted between shutdown and transport
- 365 days, 100 days, or 5 days
96
- Initial temperature dictated by cool down time Temperatures are much lower than operation
- Relatively low rate of temperature increase High thermal capacity of core components Average assembly temperature Decreasing cool down time Decreasing cool down time Decay heat Scenario 3:
Transport Accident
97 Scenario 3:
Transport Accident BA class 365 days 100 days 5 days With package and pressure boundary damage, gaseous fission product may release into environment 365 days 100 days 5 days XE class Cool down time does not impact fission product release, given severity of accident
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 Energy's National Nuclear Security Administration under contract DE-NA0003525.
MELCOR Summary
99 MELCOR capabilities were demonstrated New TRISO fueled, horizontal HPR model developed Demonstration of generalized framework Application with ORIGEN module in MELCOR Key insights Low power density significantly mitigates accident consequences I-135/Xe-135 equilibrium ratio during operation impacts Xenon feedback Considering Xenon feedback critical to capturing long term transient behavior Transport accident not expected to cause additional fission product release from fuel Future work Expand on physics models and capabilities within generalized architecture MELCOR HPMR Fuel Cycle Demo Summary Reactor room vents
MELCOR Validation and Verification Basis 100
SCALE Benchmarking and Validation Activities 101 101 HTGRs MSRs SFRs SCALE Validation in Four Major Areas (Criticality Safety, Radiation Shielding, Reactor Physics, and Spent Fuel Inventory)
10 2
New modeling capabilities have been added to SCALE & MELCOR for a heat pipe microreactor Leveraged capabilities developed in previous non-LWR demonstration workshops New modeling demonstrated with multiple scenarios from various stages of the fuel cycle to develop an understanding of microreactor accident behavior and overall plant response
- This and previous workshops have positioned the agency to readily support the review of microreactor designs and guide the reviews significant issues Closing Remarks
List of Acronyms AC - Actinides FHR - Fluoride Salt Cooled High Temperature Reactor MSBR - Molten Salt Breeder Reactor ACRS - Advisory Committee for Reactor Safeguards FP - Fission Products MSR - Molten Salt Reactor BeO - Beryllium Oxide GWd - Gigawatt-days MSRE - Molten Salt Reactor Experiment BOC - Beginning of Cycle Gy - Gray MSTDB - Molten Salt Thermal Properties Database BOL - Beginning of Life HALEU - High Assay Low Enriched Uranium MTIHM - Metric Ton of Initial Heavy Metal CD - control drum HP - Heat Pipe MWth - Megawatt Thermal CE - Continuous Energy HPMR - Heat Pipe Microreactor OGS - Off Gas System CFR - Code of Federal Regulations HPR - Heat Pipe Reactor OPyC - Outer Pyrolytic Carbide Ci - Curie HS - Heat Structure Pa - Pascal CP - Construction Permit HTGR - High Temperature Gas Cooled Reactor pcm - per cent mille CSAS - Criticality Safety Analysis Sequences IAP - Implementation Action Plan ppm - parts per million CV - Control Volume INL - Idaho National Laboratory PWR - Pressurized Water Reactor DOE - Department of Energy IPS - Iron Pipe Size SFR - Sodium Fast Reactor EFPD - Effective Full Power Days IPyC - Inner Pyrolytic Carbide SiC - Silicon Carbide EFPY - Effective Full Power Years LANL - Los Alamos National Laboratories Sv - Sievert ENDF - Evaluated Nuclear Data File LT - Light Elements TRISO - Tri-structural Isotropic EOC - End of Cycle LWR - Light Water Reactor UCB - University of California Berkely EOL - End of Life MAVRIC - Monaco with Automated Variance Reduction using Importance Calculations UCO - Uranium Oxycarbide EOS - Equation of State MeV - Mega Electronvolt V&V - Verification and Validation FA - Fuel Assembly MG - Multigroup