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Recycled Nuclear Fuel Cost Calculator

(last updated 23 Mar 2020)
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Calculates the material balance and cost of nuclear fuel (for light-water reactors) produced from

enriched natural uranium (U_nat),
re-enriched depleted uranium (U_dep),
enriched uranium recycled from spent fuel (U_rep),
plutonium recycled from spent fuel (MOX), and
uranium blended down from highly enriched uranium (HEU).

Enter a value into any one field of the Material Balance and click the "Calculate" button below. HELP
"t" = metric tonne. "HM" = heavy metal. "Pu_f" = fissile plutonium (Pu-239 and Pu-241). "wt-%" = weight-percent.
"RB" = Raw Blendstock. "FB" = Final Blendstock.

Material Balance
	U_nat	U_dep	U_rep	MOX	HEU
Source Material	t U₃O₈ (= t U_nat)	t U₃O₈ (= t U_dep)	t UO₃ (= t U_rep)	kg Pu (= kg Pu_f) t UO₂ (= t U)	kg UO₃ HEU (= kg U) t U₃O₈ RB (= t U)
Conver- sion	t UF₆ (nat)	t UF₆ (dep)	t UF₆ (rep)		t UF₆ HEU t UF₆ RB
Re-Enrich- ment		SWU t UF₆ (nat-equiv.)
Enrich- ment	SWU t UF₆ (enriched)	SWU t UF₆ (enriched)	SWU t UF₆ (rep,enriched to equiv.)		SWU t UF₆ FB
Fuel Fab.	t UO₂ (= t U)	t UO₂ (= t U)	t UO₂ (= t U)	t MOX (= t HM)	t UO₂ (= t U)
Power Plant	GWh_e ( = GWa_e)

Cost Summary (in Thousand US$)

U_nat U_dep U_rep MOX HEU

Source Material

Conversion

Re-Enrichment

Enrichment

Fuel Fabrication

Total

HELP

Cost Summary Chart

Cost Parameters

U_nat U_dep U_rep MOX HEU

Source Material
$ / lb U₃O₈ $ / kg U $ / kg U $ / kg Pu
$ / kg U $ / kg U HEU
$ / kg U RB

Conversion $ / kg U $ / kg U $ / kg U $ / kg U HEU

Re-Enrichment $ / SWU

Enrichment $ / SWU $ / SWU

Fuel Fabrication $ / kg U $ / kg U $ / kg HM $ / kg U

Assay Summary

U_nat U_dep U_rep MOX HEU

[wt-% U-235] [wt-% U-235] [wt-% U-235] [wt-% U-236] [wt-% U-235] [wt-% Pu_f] [wt-% U-235] [wt-% U-236]

Feed (re-enr.)
(enr.) (HEU)
(RB)
(FB) (HEU)

Fuel

Tails (re-enr.)
(enr.)

(for display only)

Process Parameters

Source
Material U_nat U_nat enrichment to initial enrichment (see Power Plant)
U_nat tails assay: wt-% U-235
U_dep Step 1: U_dep re-enrichment to natural equivalent (0.711 wt-% U-235)
U_dep feed assay: wt-% U-235 · tails assay: wt-% U-235
U_dep supplied as: U₃O₈, UF₆
Step 2: further enrichment to initial enrichment (see Power Plant); tails assay: see U_nat
U_rep U_rep reprocessed after 5 year storage of spent fuel after unload from reactor
U_rep enrichment to initial enrichment equivalent (see Power Plant); tails assay: 0.2 wt-% U-235
MOX Pu_f equiv. of U-235: g Pu_f per g U-235
Pu_f conc. in total Pu: wt-% Pu_f
MOX U component assay: wt-% U-235

HEU HEU assay: wt-% U-235 · wt-% U-236

Blendstock sample:
DU = depleted uranium · Unat = natural uranium · SEU = slightly enriched uranium

Final Blendstock (FB) assay: wt-% U-235

If Final Blendstock is obtained by enrichment of Raw Blendstock:
Raw Blendstock (RB) assay: wt-% U-235 · Tails assay: wt-% U-235

Raw Blendstock (RB) supplied as: U₃O₈, UF₆

Factor for U-236 effect: excess wt-% U-235 per wt-% U-236

Conversion Losses: %

Fuel Fabrication Losses: %

Power Plant burnup · initial enrichment:

Efficiency: %

Process Parameters
Source Material	U_nat	U_nat enrichment to initial enrichment (see Power Plant) U_nat tails assay: wt-% U-235
U_dep	Step 1: U_dep re-enrichment to natural equivalent (0.711 wt-% U-235) U_dep feed assay: wt-% U-235 · tails assay: wt-% U-235 U_dep supplied as: U₃O₈, UF₆ Step 2: further enrichment to initial enrichment (see Power Plant); tails assay: see U_nat
U_rep	U_rep reprocessed after 5 year storage of spent fuel after unload from reactor U_rep enrichment to initial enrichment equivalent (see Power Plant); tails assay: 0.2 wt-% U-235
MOX	Pu_f equiv. of U-235: g Pu_f per g U-235 Pu_f conc. in total Pu: wt-% Pu_f MOX U component assay: wt-% U-235
HEU	HEU assay: wt-% U-235 · wt-% U-236 Blendstock sample: DU = depleted uranium · Unat = natural uranium · SEU = slightly enriched uranium Final Blendstock (FB) assay: wt-% U-235 If Final Blendstock is obtained by enrichment of Raw Blendstock: Raw Blendstock (RB) assay: wt-% U-235 · Tails assay: wt-% U-235 Raw Blendstock (RB) supplied as: U₃O₈, UF₆ Factor for U-236 effect: excess wt-% U-235 per wt-% U-236
Conversion	Losses: %
Fuel Fabrication	Losses: %
Power Plant	burnup · initial enrichment: Efficiency: %

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