1. ULB 2009-2010 Nuclear Fuel Cycle Nuclear Fuel reprocessing Sellafield - UK

2. ULB 2009-2010 Nuclear Fuel Cycle Nuclear fuel reprocessing 1.Why reprocess? 2. Basic principles 3.Description of PUREX process 4.Industrial status

3. ULB 2009-2010 Nuclear Fuel Cycle Reprocessing objectives  Recycling of fissile materials (U, Pu),  Reduction of U needs)  Reduction of high level waste volumes  Reduction of radiotoxicity and heat from the waste

4. ULB 2009-2010 Nuclear Fuel Cycle The Reprocessing-Recycling Note: message AREVA

5. ULB 2009-2010 Nuclear Fuel Cycle Fissile materials recycling Spent UOX fuel (33 GWj/t, cooling 3 years)

6. ULB 2009-2010 Nuclear Fuel Cycle Spent fuel composition

7. ULB 2009-2010 Nuclear Fuel Cycle La radiotoxicité des déchets

8. ULB 2009-2010 Nuclear Fuel Cycle Arguments against reprocessing  Technological difficulty and large investments  Large, generally export, reprocessing costs  Accumulation of Pu: recycling need  Nuclear proliferation need  Transports of nuclear materials

9. ULB 2009-2010 Nuclear Fuel Cycle Le retraitement du combustible irradié 1.Why reprocess? 2. Basic principles 3.Description of PUREX process 4.Industrial status

10. ULB 2009-2010 Nuclear Fuel Cycle Reprocessing functions 1.Separation from spent fuel of U, Pu, and Fission Products (FP)+ Minor Actinides (MA) 2.Purification of U and Pu, to be re-used 3.Concentration of FP + MA for final geological disposal

11. ULB 2009-2010 Nuclear Fuel Cycle  Developed by Oak Ridge National laboratory (ORNL) and Knolls Atomic Power Laboratory (KAPL) from 1949 to 1960  Solvent extraction based on TBP  Targeted for separation of U and Pu  Used on an industrial scale in Savannah River & Hanford (USA, past), La Hague (F), Sellafield (UK), Rokkashamura (J) PUREX: Plutonium Uranium Refining by EXtraction

12. ULB 2009-2010 Nuclear Fuel Cycle UP3 La Hague plant

13. ULB 2009-2010 Nuclear Fuel Cycle Nitric acid  Due to various oxidation states of N, allows the change of actinides valences  Not too corrosive, formation of soluble metal nitrates  Stability in nitric acid medium: UVI NpV and NpVI PuIV and PuVI AmIII  Recycling of vapours in nitric acid (2NO+O 2 N 2 O 4 +H 2 O)

14. ULB 2009-2010 Nuclear Fuel Cycle U chemical properties  Electronic configuration: [Rn]5f 3 6d 1 7s 2  6 extractible valence electrons: U metal oxidises easily in humid or hot air  Complex chemistry (5f electrons): oxidation levels III to VI  Level VI most stable (uranyle UO 2 2+ in solution)  Uranyle nitrate solubility in various organic compounds

15. ULB 2009-2010 Nuclear Fuel Cycle Plutonium chemical properties  Electronic configuration: [Rn]5f 6 6d 0 7s 2  Reuslts from neutronic irradiation of U  Mix of several isotopes: 238, 239, 240, 241, 242  Oxydation levels III to VII  Levels III and IV in industrial processes  Final reprocessing product: PuO 2

16. ULB 2009-2010 Nuclear Fuel Cycle Physico-chemical aspects (1)  Fuel rods/assemblies mechanical shearing (3-4 cm slices)  Fuel dissolution in boiling nitric acid (2h) UO 2 + 4HNO 3 → UO 2 (NO 3 ) 2 + 2NO 2 + 2H 2 O UO 2 + 3HNO 3 → UO 2 (NO 3 ) 2 + 0,5NO 2 + 0,5 NO + 1,5H 2 O Nitrates: Pu (NO 3 ) 4, PF (NO 3 ) 3, MA(NO 3 ) 3  Structural materials conditioning (high activity solid waste)  Nitrous vapours treatment  Volatile and gaseous FP treatment

17. ULB 2009-2010 Nuclear Fuel Cycle Physico-chemical aspects (2)  TBP: organic compound forming complexes with metal (M) nitrates, not soluble in water  M aq x- + xNO 3aq - + y TBPorg [M(NO 3 ) x y TBP]org Formation of complexe controled by concentration in ions NO 3 -  Increase NO 3 - favours extraction of M in organic phase  Decrease NO 3 - favours re-extraction of M in aqueous phase

18. ULB 2009-2010 Nuclear Fuel Cycle (C 4 H 9 ) 3 PO 4 or PO(OC 4 H 9 ) 3  Low solubility in aqueous phase  Affinity for U VI and Pu IV (selectivity)  Good chemical resistance to radiolysis  Density: 0.973 gcm -3 ; if 30% diluted: 0.83 gcm-3 TBP = tri-butyl phosphate Twin free oxigen electrons

19. ULB 2009-2010 Nuclear Fuel Cycle UO 2 + 2 NO 3 + 2TBP = UO 2 (NO 3 ) 2. 2TBP The distribution coefficient (coéfficient de partage) D is the ratio of the concentration in the aqueous and organic phase: Distribution coefficient

20. ULB 2009-2010 Nuclear Fuel Cycle Distribution coefficient

21. ULB 2009-2010 Nuclear Fuel Cycle Extraction ability ClassAbility to form complexes with TBP Extraction ability A) UO 2+, PuO 2 2+, Pu 4+, U 4+, Zr 4+, Ce 4+, RuNO 2 3+ Relatively highVery good to good B) Pu 3+, Y 3+, Ce 3+ LowLow to very low C) Other FPsVery low to nilAlmost nil

22. ULB 2009-2010 Nuclear Fuel Cycle TBP HNO 3 Spent fuel U Pu Fission products Minor actinides Xe, Kr, I 2 PUREX Principle TBP en solution dans hydrocarbure (30%) Emulsion Transfert de matières Mélange Décantation

23. ULB 2009-2010 Nuclear Fuel Cycle Separation U - Pu  Pu 4+ extracted with U (class A)  Pu 3+ class B : low ability to form complexes  Mixing of organic phase with aqueous solution, containing a selective Pu reductor (concentration NO 3 - must be sufficient to keep U in organic phase)  During emulsion of the phases, Pu is reducted and goes in the aqueous phase -

24. ULB 2009-2010 Nuclear Fuel Cycle Purification U and Pu  Impureties: FPs of class A  Extraction ability lower than U and Pu, depending on [U] and [nitric acid]  High [U]: mitigates FPextraction  High acidity: decreases Ru extraction increases Zr, Sr extraction  Successive washing of organic phase  Concentration NO 3 - variable, but sufficient pour hinder the re- extraction of U and Pu!

25. ULB 2009-2010 Nuclear Fuel Cycle TBP separation basic principles Sélectivity of TBP (UVI and PUIV) Importance of acidity: to extract UVI and PuIV: 2-3 mol/l To de-extract UVI: <0,02 mol/l Separation U-Pu: reduction PuIV to PuIII Separation U-Np: adjustment of the Np oxidation state to NpV Am is not extracted by TBP

26. ULB 2009-2010 Nuclear Fuel Cycle Le retraitement du combustible irradié 1.Why reprocess? 2. Basic principles 3.Description of PUREX process 4.Industrial status

27. ULB 2009-2010 Nuclear Fuel Cycle Shearing Dissolution Clearing Extraction Purification Spent fuel U Pu Structural elements Hulls Hulls Fission products & MA Insoluble residues Vitrification Gases Gases Atmospheric or sea release PUREX: Plutonium URanium EXtraction

28. ULB 2009-2010 Nuclear Fuel Cycle http://www.ricin.com/nuke/bg/lahague.html The La Hague reprocessing scheme

29. ULB 2009-2010 Nuclear Fuel Cycle 29 Spent fuel assemblies storage pool at Sellafield (UK)

30. ULB 2009-2010 Nuclear Fuel Cycle Shearing of cladding

31. ULB 2009-2010 Nuclear Fuel Cycle Rotatif Dissolver

32. ULB 2009-2010 Nuclear Fuel Cycle Caractéeristics of the dissolution solution Composition: U: 200 – 250 g/L Pu: 2 – 3 g/L FP: 80% of inventory MA: 100% Specific activity : 7,4 TBq/L (200Ci/L) Nitric acidity : 3 – 4 M Oxidation state of oxides: VI, PuIV, NpV, AmIII, CmIII

33. ULB 2009-2010 Nuclear Fuel Cycle Extraction cycles in a reprocessing plant (example) 1.Decontamination – separation cycle  M Extraction in organic phase  Acid washing of the organic phase  Pu Separation (reducing re-extraction)  U Re-extraction in aqueous phase 2.U purification cycles (2x)  New U extraction in organic phase  Washing  U re-extraction in aqueous phase 3.Pu purification cycles (2x)  Solution oxidation (Pu 4+ )  New Pu extraction in organic phase  Pu re-extraction in reducing aqueous phase

34. ULB 2009-2010 Nuclear Fuel Cycle Feed (aq) Product (org) Waste (aq) Fresh solvent (org) Fresh solvent Aqueous feed Loaded solvent meets most concentrated aqueous solution Fresh solvent meets depleted aqueous solution Counter current: maximising loading & extraction

35. ULB 2009-2010 Nuclear Fuel Cycle Feed(aq)Product(org) Waste(aq) cfcfcfcf cpcpcpcp cwcw Fresh solvent (org) c = 0 1 i n Multi-stage extraction

36. ULB 2009-2010 Nuclear Fuel Cycle Solvent extraction devices

37. ULB 2009-2010 Nuclear Fuel Cycle Solvent extraction devices

38. ULB 2009-2010 Nuclear Fuel Cycle Laboratory scale centrifugal contactors (ITU)

39. ULB 2009-2010 Nuclear Fuel Cycle Pulsed Column

40. ULB 2009-2010 Nuclear Fuel Cycle Solvent extraction devices

41. ULB 2009-2010 Nuclear Fuel Cycle Recovery rate and decontamination factor Residual materials recovery rate: Pu:99,88% Decontamination factor: Impureties in inlet product divided by impureties in outlet product β, γ impurities: U: 1,5 10 6 ; Pu: 7 10 7 Separation factor U-Pu: 10 6

42. ULB 2009-2010 Nuclear Fuel Cycle 42 Technological constraints of reprocessing High activities Heat release Under-criticity to be guaranteed, verifications Corrosion resistance (stainless steels, zirconium) Maintenance of equipement Controls of materials fluxes

43. ULB 2009-2010 Nuclear Fuel Cycle U and Pu conditioning  Aqueous solution of Uranyl nitrate [UO 2 (NO 3 ) 2 ] at 250 – 300 g U / l  Denitration and transformation into UO 3 or UO 2 (fabrication plant)  Aqueous solution of Pu nitrate: [Pu (NO 3 ) 2 ] at 50-150 g Pu / l  Oxidation of Pu in Pu 4+, mixing to oxalic acid which precipitates Pu as oxalate  Calcination and storage of PuO 2 or transport to MOX plant

44. ULB 2009-2010 Nuclear Fuel Cycle Plutonium Conversion : calcination

45. ULB 2009-2010 Nuclear Fuel Cycle  High decontamination factors  High selectivity for U and Pu  Low cost  Easy scale up  Room temperature process  Radiolytic degradation of organic phase  TBP not incinerable yielding solid radioactive waste  Some fission products are not (fully) soluble (Zr, noble metals particles)  Pure plutonium produced Advantages and disadvantages of PUREX

46. ULB 2009-2010 Nuclear Fuel Cycle  Bitumen: e.g. for residues from evaporation or spent organic ion exchangers  Cement: for low radioactive waste  Glass: for high level liquid waste  Ceramics: alternatives for HLLW (not industrial) Waste forms

47. ULB 2009-2010 Nuclear Fuel Cycle  Borosilicate glass matrix  HLW concentrate is calcined  Mixed with glass frit and heated at 1100 o C  Liquid poured in a stainless steel canister  Canister is welded shut Vitrification of HLW

48. ULB 2009-2010 Nuclear Fuel Cycle  Silica is the main glass- forming component  Boron oxide reduces thermal expansion and improves chemical durability Vitrification of HLW

49. ULB 2009-2010 Nuclear Fuel Cycle Vitrification of HLW

50. ULB 2009-2010 Nuclear Fuel Cycle Waste treatment

51. ULB 2009-2010 Nuclear Fuel Cycle Le retraitement du combustible irradié 1.Why reprocess? 2. Basic principles 3.Description of PUREX process 4.Industrial status

52. ULB 2009-2010 Nuclear Fuel Cycle 52 Reprocessing capacities in the world LWR fuel:France, La Hague1700 UK, Sellafield (THORP)900 Russia, Ozersk (Mayak)400 Japan14 total approx3000 Other nuclear fuels: UK, Sellafield1500 India275 total approx1750 Total civil capacity 4750 NEA 2004

53. ULB 2009-2010 Nuclear Fuel Cycle 53 Rokkasho-Mura (Japan)

54. ULB 2009-2010 Nuclear Fuel Cycle AREVA La Hague Reprocessing Plants

55. ULB 2009-2010 Nuclear Fuel Cycle 55 UP3 plant in La Hague

56. ULB 2009-2010 Nuclear Fuel Cycle 56 Marcoule R&D

57. ULB 2009-2010 Nuclear Fuel Cycle Conclusion  Reprocessing: strategic option  based on nitric dissolution, séparation by organic extraction  Reprocessing-Recycling strategy, in LWRs, but preferably in fast reactors  Technical and commercial success  3 main sites: FR, UK, JP Thank you for your attention!