Development of a MOX equivalence Python code package for ANICCA

. The basis of the MOX (Mixed OXide) energy equivalence principle is keeping the in-core fuel management characteristics (cycle length, feed size, etc.) of a nuclear reactor unchanged when replacing UOX (Uranium OXide) fuel assemblies by MOX. If the effect of the loading pattern is neglected, such an equivalence is obtained by tuning the Pu content in the MOX fuel, while considering the specific Pu isotopic vector at the time of the core reload to obtain a crossing of the reactivity curves of UOX and MOX at the end-of-cycle core average burnup. It is proposed in this work to extend the fuel cycle analysis tool ANICCA (Advanced Nuclear Inventory Cycle Code) with a MOX equivalence Python code package, which automatically governs the supply and demand of Pu vector isotopes required to obtain MOX equivalence. This code package can determine the reactivity evolution for any given Pu vector by means of a multidimensional interpolation on a directive grid of pre-calculated data tables generated by WIMS10, covering the physically accessible Pu vector space. A fuel cycle scenario will be assessed for a representative evolution of the Pu vector inventory available in spent UOX fuel as a demonstration case, defining the interim fuel storage building dimensional requirements for different reprocessing strategies.

o Reprocessing strategy for spent UOX fuel will be defining parameter in evolution of spent fuel inventory: FIFO (First In, First Out) or LIFO (Last In, First Out)?
o Identify possible need for interim storage buildings and associated capacity dimensioning o Analysis may become very complex as difference in origin (different PWRs) of spent fuel, irradiation history (burnup), and cooling time all introduce additional dispersion to Pu vector  • ANICCA = Advanced Nuclear Inventory Cycle Code: ANICCA • Fuel cycle analysis tool to monitor flow of nuclear material between facilities • Python code developed at SCK CEN (Belgium) • Flexible/modular code allowing for easy modification of scenarios but also for further code development Equivalence target in following ICFM:

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• Industrial MELOX process limited <12% Pu max, or <10.6% average when accounting for radial zoning • FIFO: almost sensitive to delay "hot fuel is a volatile resource" 241 Pu "extraction of spent UOX for MOX fabrication" • LIFO strategies are more beneficial in terms of heat load removal even though inventory remains higher than FIFO at all "decay heat in spent UOX stockpile and Pu reprocessing facility" • Scenario analysis for capacity dimensioning of interim spent fuel storage buildings: o Gradual phase-out of all but two PWRs between t 0 and t 0 + 6 years o One remaining PWR continues UOX operation until t 0 + 13.5 years o Other remaining PWR switches at t 0 to: • ¼ MOX -FIFO -1.5 yrs delay (reprocessing → loading) • ¼ MOX -LIFO -1.5 yrs delay (reprocessing → loading) • Full UOX core (as before) • On-site spent fuel inventory growth can be reduced to +18 à 20% instead of +36%!

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Introduction Tools and methods o ANICCA -Advanced Nuclear Inventory Cycle Code: ANICCA o Directive Pu vector mesh generation o Linear Reactivity Model & MOX energy equivalence principle • Case study: fuel reprocessing of representative irradiated fuel stock o Fuel cycle scenario description o Results and discussion • Conclusion 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) PUBLIC • Situation = decision for spent UOX fuel reprocessing is taken after long period of once through operating mode:

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Mid-and long-term cycle calculations: o Nuclear power plant fleet management o Waste characterization o Reprocessing of spent fuel o … 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) PUBLIC • Dispersion of average Pu isotopy of MOX batch is mainly due to following (physical) processes: Fuel assemblies with different burnups, enrichments and design (e.g., 8, 12 and 14 ft assemblies) o Radioactive decay due to cooling time of fuel assembly o Radioactive decay due to delay between reprocessing and loading of fuel in core • Need to go beyond simplified equivalence model (with fixed weighting factors) depends on in-core fuel management specificities (cycle length, feed size, etc.): o Neutronic calculations required for every modification to re-determine weighting factors o Not very flexible for use in realistic (variable or perturbed) fuel cycle scenarios in ANICCA of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) PUBLIC • Build a multi-dimensional reactivity mesh for all realistically achievable: o Pu vectors ( 238 Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu, 241 Am) o Discharge burnups (0 -64 GWd/tU) o Pu fractions (6% -8% -10% -12%) • Based on empirical correlations: o Typical reference Pu vector as starting point: 21 yrs cooling time + 1 yr between reprocessing and core loading +  between [70%-100%] o perturbations based on realistic Pu vector data • ~3000 WIMS10 calculations (◼) to cover physically accessible Pu vector space and Pu fractions per assembly 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) 8  is inversely proportional to assembly burnup of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) PUBLIC • MOX equivalence Python code package for ANICCA: returns reactivity evolution for any given Pu vector covering Pu fractions (6% 8% -10% -12%) and discharge burnups (0 -64 GWd/tU) by means of interpolation on this directive Pu vector mesh 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) PUBLIC • Linear Reactivity Model (LRM) = bi-linear equation providing reactivity () as function of burnup () and U5 enrichment / Pu content () with 4 calibrated parameters:  =  0 +  *  +  *  +  *  *  • Determine required Pu content for given Pu vector and in-core fuel management requirements: ① Reactivity evolution of UOX given by Linear Reactivity Model (LRM): reactivity UOX@EOC = f(EOC burnup, initial U235 enrichment) ② Request equivalence of MOX with UOX fuel at EOC core average burnup: reactivity curves need to cross over at average EOC core burnup ③ Inverse operation on directive Pu vector reactivity mesh: Pu content = f(reactivity UOX@EOC, EOC burnup, Pu vector) of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) 11 EOC core average burnup • Illustrative application based on and realistic spent fuel stock: o Huge dispersion burnups, enrichments and cooling times o Trend to increase burnup in more recently unloaded assemblies • of 4 scenarios (with or 12 yrs delay): o FIFO = First In, First Out, "Cold first": oldest assemblies are reprocessed first o LIFO = Last In, First Out, or "Hot first": newest assemblies are reprocessed first 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021reduced 235 U support enrichment in burnable poison rods "just in time" 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) 13

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LIFO: reduced Pu requirements if MOX fuel is loaded shortly after Pu reprocessing 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) 14 "required Pu fraction to obtain energy equivalence" t 0 = assumed first reprocessing, then every 1.5 yrs 16 MOX assemblies are fabricated from irradiated fuel stock (=1/4 of feed size) • LIFO (1.5 yrs delay): less demands on reprocessing effort because less spent UOX fuel needed for fabrication, but slows down emptying of existing spent fuel stockpile • FIFO (1.5 yrs delay) + FIFO/LIFO (12 yrs delay) are rather similar 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) 15 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) PUBLIC • Fuel cycle analysis tool ANICCA (SCK CEN) has been extended with a MOX equivalence Python code package (Tractebel Engie): online calculation of Pu content requirements in MOX fuel fabrication to obtain energy equivalence for different types of in-core fuel management • Best choice of scenario depends on specific needs: o LIFO = Last In, First Out, or "Hot first": much less spent UOX to reprocess for same energetic content in MOX fuel = reduced reprocessing effort o FIFO = First In, First Out, or "Cold first": accelerated emptying of spent fuel pools = reduced storage facility capacity requirements o Exercise needs to done for each specific case as results depend on storage constraints, in-core fuel management, equivalence objectives, acceptable MOX fraction, … very attractive to think about and optimise it!01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021)

238 Pu + 241 Am
During storage: 241 Pu decays into 241 Am • Scope = extend ANICCA (Advanced Nuclear Inventory Cycle Code), a fuel cycle analysis tool developed at SCK CEN (Belgium), with MOX equivalence Python code package: o Determine reactivity evolution for any given Pu vector by means of multidimensional interpolation on mesh of pre-calculated data tables generated by WIMS10, thereby covering physically accessible Pu vector space o Perform online calculation of Pu content requirements in MOX fuel fabrication for a given fuel cycle scenario to obtain energy equivalence 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021) PUBLIC • Neutronics: o o Reprocessing: 241 Am is eliminated o After reprocessing: new 241 Am accumulation 241 Am, 240 Pu and 242 Pu are neutron absorbers • Storage and fabrication: o Residual heat: • Radiation protection: o 238 Pu, 241 Am (α, n) on 17 O & 18 O o (weak γ by 241 Am) 01/07/2021 Development of a MOX equivalence Python code package for ANICCA -5th Technical Workshop on Fuel Cycle Simulation (TWoFCS 2021)