ML26037A058
| ML26037A058 | |
| Person / Time | |
|---|---|
| Issue date: | 02/06/2026 |
| From: | Nathanael Hudson NRC/RES/DSA/FSCB |
| To: | |
| References | |
| Download: ML26037A058 (0) | |
Text
Fission Data for Regulatory Needs Workshop for Applied Nuclear Data Activities (WANDA) 2026 February 11, 2026 Nathanael Hudson US Nuclear Regulatory Commission Office of Nuclear Regulatory Research Division of Systems Analysis
- Lattice/Nodal Methods (10 CFR Part 50/Part 52)
- Tech Specs Limits, FSAR, and SE based on reactor physics calculations
- Detailed neutron transport and depletion calculations at lattice level are condensed in energy and homogenized in space
- Combined with nodal diffusion methods at the core level homogenization x
y z
0
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(
z y
x h
h h
2/11/2026 2
Fission Data - Reactor Physics
- Fuel Facilities (10 CFR Part 70); Independent Storage of Spent Fuel (Part 72); Packaging and Transportation of Radioactive Material (Part 71)
- Structural, thermal, criticality safety, radiation shielding analysis, (confinement/retrievability-Part 72)
ORANOs DN30-X for UF6 up to 20 wt.%
U-235 NACs Optimus-L for TRISO compacts
- Detailed models of critical configurations in facilities and packages transporting fresh and depleted fuel to calculate neutron multiplication factor (keff) and uncertainty (keff < 1 with margin) 2/11/2026 3
Fission Data - Criticality Safety
- Part 70, Part 71, Part 72, Part 73
- 10 CFR Part 73.37 -- Physical protection of irradiated fuel during transit - 100 rad/h at 1m distance from any accessible surfaces without intervening shielding
- Depletion to develop inventory for neutron and primary and secondary gamma sources to approximate dose Dose map from Microreactor Fuel Cycle Workshop (March 26, 2025)
- Non-LWR source term public workshop videos, slides, reports at the NRC
- SCALE/MELCOR demonstrations 2/11/2026 4
Fission Data - Shielding/Dose
- Foundation for calculations to demonstrate design and safety
- Essential for predicting neutron multiplication and power Fission Cross Sections -
Probability of fission for each fissile isotope at different neutron energies Multiplication factor (keff)
Integral power Node-wise flux/power shapes (r,z)
Pin-wise power Reactivity feedback coefficients (fuel temperature, moderator density, void coefficients, etc.)
Shutdown Margin - Reactivity control mechanisms (boron, control rods) are sufficiently large to overcome postulated power rise Neutron energy distribution, impacting fuel utilization and power distribution 2/11/2026 5
Fission Cross Sections
Prompt Prompt Spectrum (E)
Delayed Group fractions Decay constants Delayed neutron fractions Precursor yields Reactivity coefficients Reactivity control (reactor response)
Energy release/fission Power normalization LOCA PCT Reactivity Insertion Accidents Stability Analysis ATWS Control Rod Eject Reactor Period Analysis 2/11/2026 6
Kinetics Inputs - Core Transient
Fission Product Yields Decay heat and isotopic inventories -- important for spent fuel evaluations and source term during reactor severe accidents Burnup credit - accurate isotopic inventories determine how much reactivity credit can be taken for spent fuel Radiotoxicity and waste classification - depend on long-lived fission products Shielding calculations - require correct gamma-emitter inventories & neutron source terms Fuel performance (e.g., swelling, gas release) depends on noble gas yields
- Fission product yields (FPYs) define the distribution of nuclides produced in fission Errors propagate throughout any calculation that depends on depletion (fuel cycle, source term, transient analysis) 2/11/2026 7
Fission Product Yields
- Energy-dependent FPYs vary with incident neutron energy (thermal, epithermal, fast) and interpolated from discrete points (ENDF/B),
depending on nuclide Transport solver (TRITON, Polaris, KENO, or Shift) computes flux Depletion module (ORIGEN) uses FPYs and decay data to develop isotopics Sources of Uncertainty Sparse experimental data above thermal (several non-LWR designs characterized by epithermal/fast spectra)
Interpolations would benefit from a denser incident neutron energy grid for key actinides within current formats Continuous energy representation of tabulations would be most ideal Fast-spectrum yields for Pu isotopes and U-238 have high uncertainties More integral validations Sparse covariance data Distorted uncertainties Mismatches between independent and cumulative yields 2/11/2026 8
Energy-Dependent Fission Product Yields in SCALE
- Regulatory Review Expectations:
- Validation against critical experiments
- Bias and uncertainty quantifications
- Conservative margins
- Poor fission data sources:
- Larger safety margins
- More restrictive operating limits
- Reduced fuel utilizations
- Difficulty quantifying advanced reactor designs 2/11/2026 9
Regulatory and Licensing Implications