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@@ -1,3 +1,31 @@
+- title: "Burning fuel for cheap! Transport-independent depletion in OpenMC"
+  presenter: Oleksandr Yardas
+  event:  (SciPy 2025)
+  type: Technical Presentation
+  place: Tacoma, WA
+  month: July
+  day: 10 
+  year: 2025
+  pic: '/img/pres/2025-scipy.png'
+  slides: '/pres/2025-yardas-scipy.pdf'
+  description: "This presentation details a new method for running depletion
+  simulations independently of neutron transport in OpenMC.
+  Transport-independent depletion uses pre-computed static multigroup cross
+  sections and fluxes to calculate reaction rates for OpenMC's depletion
+  matrix solver. This accelerates the depletion calculation, but removes the
+  spatial coupling between depletion and neutron transport. Using this
+  method, concentration errors for low-abundance nuclides at longer (30-day)
+  time steps exhibit large negative initial concentration the becomes more
+  positive with time due to overestimation of nuclide production stemming
+  from the lack of spatial coupling to neutron transport. For ten 3-day time
+  steps, fission product concentration errors are all under 3\%. Actinide
+  concentration errors range from 10-15\% for Am and Cm, 5-7\% for Pu and Np,
+  and 2\% and less for U. Surprisingly, the numbers are similar for 30-day
+  time steps. These results demonstrate the potential of this new method with
+  moderate accuracy and extraordinary time savings for low and medium
+  fidelity simulations. Concentration error characterization on larger models
+  remains an open area of interest."
+
 - title: "A Hybrid SN-Diffusion Method for Molten Salt Reactor Control Rod Modeling"
   presenter: Sun Myung Park
   event: International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
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