ML25261A334

From kanterella
Jump to navigation Jump to search
RES Slides Sept 2025 Workshop (ML25261A334)
ML25261A334
Person / Time
Issue date: 09/18/2025
From:
Office of Nuclear Regulatory Research
To:
References
Download: ML25261A334 (1)
v • d • e


Text

Updates to Methodology for Containment Source Term September 24, 2025 Fuel & Source Term Code Development Branch Division of Systems Analysis Office of Nuclear Regulatory Research

Part 1: Methodology for PWR Containment Aerosol Removal Factors Shawn Campbell, PhD Office of Nuclear Regulatory Research Division of Systems Analysis Fuel and Source Term Code Development Branch K. C. Wagner Lindsay Gilkey David Luxat, PhD Severe Accident Modeling and Simulation Sandia National Laboratories

Outline

  • Background and Motivation
  • Source Term Methodology
  • Revised Aerosol Removal Rates for PWR Containments 3

Background and Motivation 4

5 In-Vessel Ex-Vessel Fuel heat up Clad oxidation Core relocation Early containment failure?

Vessel Breach Late containment failure?

MCCI/FP Release Containment Source Term (ST)

Leak Rate ()

FP Inventory

()

Dose Calculation C0 =ST/ Vol Containment Leakage FP Release and Transport C t = C0 exp()

FP release = C FP removal mechanisms ()

e.g., Sprays/natural deposition Containment Source Term (ST)

Integrated Analysis (e.g., L3PRA,

SOARCA, Fukushima)

Regulatory Source Term Analysis (for DBA)

Mechanistic Modeling User Specified Simplified Modeling User Specified Simplified Modeling Source Term Methodology Part 2 Part 1

Key MELCOR Projects Supporting Best-estimate Accident Modeling 6

RADTRAD Removal rates for airborne aerosols in containment have historically been informed by NUREG/CR-6189 (1996) which evaluated the settling rate under a range of containment conditions Aerosol removal factors () can be evaluated using MELCOR in a more integrated fashion

[INL/EXT-21-61607, Rev 2]

Workflow Step 1: Establish Thermal-Hydraulic and Radionuclide Sources 7

Sources to Containment

  • Steam
  • Water
  • Fission Products Severe accident progression up to vessel breach
  • Same models and scenarios as the High Burnup Fuel Source Term Accident Sequence Analysis,

[SAND2023-1313]

o 12 Sequences: 5 for large dry (Surry) and 7 for ice condenser (Sequoyah) containments

  • Run scenarios up to the point of vessel breach
  • Capture all thermal-hydraulic and radionuclide sources to the containment

Workflow Step 2: Containment Analysis 8

HX Use fast-running containment model to quantify aerosol deposition

  • Transfer containment sources to a fast-running containment model to quantify aerosol deposition rates
  • Separate containment models are made for each sequence
  • Assume that safety systems are recovered just prior vessel breach to arrest the accident progression (i.e., decay heat being removed)
  • New sequence-specific containment models simulate in-vessel phase and the long-term phase End of the in-vessel accident phase

9 Parameter Description Range Distribution Type Reactor Types Large dry containment design or ice condenser design Sequoyah & Surry Uniform Sequence types Initiating events from SAND2023-01313 5 Surry sequences 7 Sequoyah sequences Uniform AMMD Aerosol mass median diameter at the release locations 0.2 to 2 micron Log-uniform Aerosol GSD Aerosol geometric standard deviation about the AMMD 0.2 to 5 Normal Aerosol shape factor Dynamic shape factor for aerosol settling, see NUREG/CR-7262.

1 to 5 75% between 1 to 2 Beta Radionuclide class mass multipliers Lower extension from NUREG/CR-6189 to reduce agglomeration.

0.5 to 1.5 Triangular m= 1 Containment wall heat transfer multiplier Uncertainty parameter per NUREG/CR-7626 & NUREG/CR-6189 1 to 2 Triangular Steam solubility multiplier Hygroscopic model parameter multiplier for CsOH, CsI, and Cs2MoO4 [kg/kg-H2O]

0 to 1 Triangular peaking at 1 kp, thermal conductivity of aerosols Thermal conductivity of aerosol material.

0.1 to 10 Log-normal with m= near 1 Non-radioactive aerosol mass multiplier Includes control rods and structural material aerosols (e.g.,

the fuel cladding) 0.5 to 2 Triangular m= 1 Identify parameters and ranges for uncertainty quantification and run evaluations o Selected method Latin Hyper-Cube Sampling: 1600 total evaluations, ~145 per sequence/case Calculate statistics on finished evaluations to establish aerosol settling rate Workflow Step 3: Uncertainty Quantification

In-Vessel Phase Effective Removal Rate For the in-vessel phase, containment removal rates are driven by high steam concentration Removal rates are higher than predicted by NUREG/CR-6189 (i.e., faster aerosol settling)

In-vessel phase results are used as the initial condition for long-term aerosol response 10 Effective Removal Rate (e ) [-]

Data Source e

NUREG/CR-6189* Gap Release 7.8e-6 NUREG/CR-6189* In-vessel Release 2.2e-5 MELCOR end of in-vessel phase, 10 percentile 1.0e-4

Long-Term Effective Removal Rate (No Sprays) 11

Long-Term Effective Removal Rate (With Sprays)

  • MELCOR results agree with median spray response
  • MELCOR Surry calculations with sprays
  • Calculated time to a 99% reduction in the airborne aerosol concentration Long-Term Effective Removal Rate (With Sprays) 13 Key Message 2 MELCOR results with sprays are consistent with NUREG/CR-5966 but with a much narrower range of uncertainty Surry MELCOR NUREG/CR-5966

Part 2: Methodology for HBU/IE Fuel Cycle Inventory Estimation Lucas Kyriazidis, Mike Call, Andy Bielen Office of Nuclear Regulatory Research Division of Systems Analysis Fuel and Source Term Code Development Branch Steve Skutnik William Wieselquist Ugur Mertyurek Nuclear Energy and Fuel Cycle Division Oak Ridge National Laboratories 14

15 In-Vessel Ex-Vessel Fuel heat up Clad oxidation Core relocation Early containment failure?

Vessel Breach Late containment failure?

MCCI/FP Release Containment Source Term (ST)

Leak Rate ()

FP Inventory

()

Dose Calculation C0 =ST/ Vol Containment Leakage FP Release and Transport C t = C0 exp()

FP release = C FP removal mechanisms ()

e.g., Sprays/natural deposition Containment Source Term (ST)

Integrated Analysis (e.g., L3PRA,

SOARCA, Fukushima)

Regulatory Source Term Analysis (for DBA)

Mechanistic Modeling User Specified Simplified Modeling User Specified Simplified Modeling Source Term Methodology Part 2

SCALEs Fuel Cycle Estimator (FCE)

  • New SCALE tool for determining various LWR core loading configurations
  • Rapid analysis of HBU/IE cores for source term generation
  • Determines end-of-cycle core-wide and batch-specific conditions and inventories 16 New easy-to-use SCALE tool for generating core-average inventories; accounting for key operational effects of enrichment, batch fractions, heavy metal loading, etc.

Generating HBU/IE LWR Inventories Use of SCALEs Fuel Cycle Estimator (FCE) to generate HBU/IE core configurations & inventories for one BWR and one PWR - Fission Product Inventory (Inv)

- FCE allows the rapid generation of representative inventories for a high burnup core Used SCALE to generate inventories with representative fuel cycle

- ORIGEN libraries based on SCALE ATF/HBU/IE Project 17 ML21088A336 ML23012A122 PWRs BWRs ML21088A354 ML23279A041

Radionuclide Grouping for Source Term Analysis 18 Radionuclide Group Member Elements Xe Group He, Ne, Ar, Kr, Xe, Rn, H, N I Group F, Cl, Br, I, At Cs Group Li, Na, K, Rb, Cs, Fr, Cu Te Group S, Se, Te, Po Ba/Sr Group Be, Mg, Ca, Sr, Ba, Ra, Es, Fm Ru Group Ru, Rh, Pd, Re, Os, Ir, Pt, Au, Ni Mo Group V, Cr, Fe, Co, Mn, Nb, Mo, Tc, Ta, W La Group Al, Sc, Y, La, Ac, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Am, Cm, Bk, Cf Ce Group Ti, Zr, Hf, Ce, Th, Pa, Np, Pu, C

PWR Cases 19 Reactor Parameters Baseline HBU HBU/IE Power (MWth) 2587 2587 2587 Initial Enrichment (%)

4.2 4.2 5.7 Cycle Length (months) 18 24 24 Fresh / Once-burned / Twice-burned 53 / 53 / 51 53 / 53 / 51 53 / 53 / 51 SCALEs Fuel Cycle Estimator Output End of cycle k 1.031 0.96 1.031 Core avg. end of cycle burnup (GWd/MTU) 38.9 53.0 52.7 Avg. assembly discharge burnup (GWd/MTU) 51.6 69.6 69.6 Objective Increase cycle length from 18 months to 24 months

PWR Inventories 20 Radionuclide Group Baseline Activity (Ci)

HBU Activity (%)

HBU/IE Activity (%)

Core-Average Burnup (GWd/MTU) 38.9 53.0 (36%)

52.7 (35%)

Xe Group 9.47E+08

-5.7%

1.3%

I Group 9.89E+08

-1.8%

-0.1%

Cs Group 1.00E+09

-4.7%

1.7%

Te Group 1.48E+09 49.9%

37.5%

Ba/Sr Group 1.33E+09

-5.5%

0.8%

Ru Group 7.68E+08 18.3%

1.5%

Mo Group 2.47E+09 3.9%

-0.2%

La Group 2.37E+09

-1.2%

1.5%

Ce Group 2.68E+09 5.2%

-2.2%

Total 1.40E+10 6.4%

4.1%

Key Message 3 SCALE Fuel Cycle Estimator provides users with the ability to quickly generate realistic LWR inventories based on specific core loading characteristics

BWR Cases 21 Objective Reduce feed batch fraction Reactor Parameters Baseline HBU/IE Power (MWth) 4016 4016 Initial Enrichment (%)

4.0 5.3 Cycle Length (months) 24 24 Fresh / Once-burned / Twice-burned 304 / 304 / 156 255 / 255 / 254 SCALEs Fuel Cycle Estimator Output End of cycle k 1.041 1.053 Core avg. end of cycle burnup (GWd/MTU) 39.6 43.5 Avg. assembly discharge burnup (GWd/MTU) 51.7 61.0

BWR Inventories 22 Radionuclide Group Baseline Activity (Ci)

HBU/IE Activity (%)

Core-Average Burnup (GWd/MTU) 39.6 43.5 (9.8%)

Xe Group 1.46E+09 4.4%

I Group 1.53E+09 1.0%

Cs Group 1.54E+09 4.2%

Te Group 3.23E+09

-5.8%

Ba/Sr Group 2.05E+09 3.7%

Ru Group 1.24E+09

-7.8%

Mo Group 3.85E+09

-2.2%

La Group 3.68E+09 2.1%

Ce Group 4.14E+09

-4.6%

Total 2.27E+10

-1.2%

23 The Fuel Cycle Estimator is designed to explore the impacts and opportunities from extended enrichment scenarios ORNL has developed a new Fuel Cycle Estimator (FCE) utility that calculates different nuclear fuel cycle configurations using basic inputs: total core power, assembly-average initial enrichment, and number of fuel assemblies per batch.

Based on multi-valued user inputs, the FCE determines key cycle characteristics such as the number of batches, assembly-average burnup at discharge, and average batch powersto enable what-if analysis of core loading scenarios.

A command-line version and a graphical user interface will be available with the SCALE 7.0 release.

24 The Fuel Cycle Estimator tool allows for exploration of different cycle configurations Users can specify multiple values for up to two parameters The Fuel Cycle Estimator tool calculates the characteristics for each permutation of cycle configurations

25 The Fuel Cycle Estimator utility relies on simple, grounded assumptions to estimate cycle characteristics

  • Linear reactivity model: After initial start-up, reactivity decreases roughly linearly with burnup
  • k as a proxy for assembly reactivity: Using lattice physics tools like Polaris, k is pre-calculated for specific configurations as a function of burnup and interpolated to problem-specific conditions
  • Batch power is proportional to reactivity: The total share of the core power produced by a batch of assemblies is proportional to k:
  • Low-leakage loading pattern: Highest-burned 20% of assemblies are loaded at the core edge and have lower power due to increased neutron leakage

= P

=1

k i

where

=1

k i = 1

26 Once an initial configuration is known, the Fuel Cycle Estimator iteratively determines the batch powers & burnups Calculate mid-cycle k

Update estimated batch powers Calculate mid-cycle burnup Initial assumption: equal batch powers, save for peripheral 20% of assemblies Calculate batch burnups from converged batch powers

27 Using multiple lattice definitions, the FCE can be extended to model BWRs NAT PSZ DOM PLE VAN N-V N-T Multiple lattice data files can be used together to represent reactivity characteristics of lattices at different moderator densities and lattice configurations (DOM, VAN)

Vanished zone (VAN):

Different lattice configuration, 70%

void Dominant zone (DOM):

30-60% void Axial blanket (NAT):

Natural uranium (no enrichment) 10% void

28 Estimation of assembly power from multiple lattice definitions relies employs a mass-weighted approach

  • For calculations employing multiple lattices, the mid-cycle k and burnup are independently calculated for each lattice in each batch
  • The batch-average k is then calculated from a mass-weighted sum of the lattice k values, i.e.:
  • Default values for weights are based on the relative fuel mass of each lattice zone (but are adjustable by the user)

=

=1

where

=1

= 1

29 The FCE provides pre-calculated lattice data for generic BWR and PWR assemblies and supports user-supplied data Users can view k and radial power peaking factor data as a function of initial enrichment and burnup Users can adjust lattice weights when using multiple lattice data files Users can import their own JSON-formatted lattice data files

30 The Fuel Cycle Estimator tool is also capable of estimating the maximum burnup location within the assembly

  • Polaris reports the distribution of relative pin powers; the maximum peaking factor is stored in the lattice data file as a function of burnup
  • The assembly axial power shape is assumed to be proportional to the lattice k relative to the same lattice j in other batches, i.e.:

where

  • The maximum pin burnup is thus:

,=

=1

= max

=1

,=

Axial peaking factor

31 Users can easily calculate batch and core-average isotopic inventories from the FCE using ORIGAMI

  • The Fuel Cycle Estimator tool can automatically generate and execute ORIGAMI inputs based on the user input (e.g., cycle length, initial enrichment) and computed cycle characteristics (batch powers)

Summary 32

  • MELCOR adds realism to accident source terms by taking an integrated approach to the long-term containment aerosol behavior o Key Message 1: Removal rates in PWRs significantly higher than NUREG/CR-6189 o Key Message 2: Results with sprays consistent with NUREG/CR-5966 but with reduced uncertainty
  • SCALE FCE tool adds realism to fission product inventory generation o Key Message 3: Quickly generate realistic inventories based on specific core loading characteristics
  • Results will be documented in reports by the end of 2025
Retrieved from "https://kanterella.com/w/index.php?title=ML25261A334&oldid=5534085"