MANAGEMENT OF HIGH LEVEL WASTE DERIVED FROM THE THORIUM NUCLEAR FUEL CYCLE

Transcription

1 Thorium Energy Conference , Shanghai, Good Hope Hotel: MANAGEMENT OF HIGH LEVEL WASTE DERIVED FROM THE THORIUM NUCLEAR FUEL CYCLE S. Yudintsev Institute of Geology of Ore Deposits, Petrography, Mineralogy and Geochemistry RAS, Moscow, Russia

2 Solutions: IV-generation high safety reactors consumed 238 U (fast neutron reactor) and 232 Th (different types). Major challenges to the nuclear energy: 1) Ore resources exhaustion. Low-cost U ores (< 80 $ per kg of U) most probably will be finished through years of exploration. 2) Safety increasing. Probability of accident must be 10-6, or 1 for 2000 years for current fleet in about 500 NPPs. Actually the value is significantly higher ( 1 per 10 years).

3 Th supplementary to U element in nuclear fuel cycle.

4 U Uranium polyisotope (3) monoelement (U) ores

5 Th Thorium monoisotope polyelement (REE, Nb) ores. Th resources may be overestimated to larger values.

6 Promising reactors IV with a Th fuel are: Very High Temperature Gas Reactor & Fast Molten Salt Reactor (ADS in future?). Existed types of reactors: SFR, AHWR.

7 Main factors influenced isolation of waste from Th nuclear cycle: Physical chemical properties of Th Phase & chemical composition of fuel Reactor type and the SNF burn up Phase & chemical composition of SNF Radionuclides contents & heat of SNF Technology of the SNF reprocessing

8 WASTE of Thorium nuclear cycle (+ tails from exploration of Thorium ores) Open cycle Closed cycle: (no reprocessing): THOREX Dry Closed cycle for: U / Pu & U / Pu & U & Th Matrix is the spent nuclear fuel (SNF) itself Waste forms: B-Si glass (FP) Glass-ceramic Ceramic (Th, MA)

9 Tails from Th deposits have lower radiotoxicity than from U ores due to difference in daughter nuclides decay: U deposits ( 238 U): 230 Th (77000 y), 226 Ra (16000 y). Th ores ( 232 Th): 228 Ra (6.7 y) & 228 Th (1.9 y), raditoxicity decreases in 20 times after 30 y.

10 Th is promising fuel for thermal & fast reactors. Fertile, so additional source of neutrons is required. Ch. Le Brun, JNM

11 Similarity and difference of Th and U cycles: to use more energetic potential of Th-based fuel high burn up is needed (>100 MWt d / kg). 232 Th has 3 times larger capture cross-section ( 238 U), 233 Pa has 11 times larger half-live period ( 239 Np).

12 Generation of 233 U and problem of 232 U.

13 Image of spent nuclear fuel Minor actinides in U / Th cycle

14 Advantage of Th cycle: low amounts of Pu and MA. Actually, MA contents strongly depend on driver fuel and in lest extent on its burn up. Separation of seed and blanket parts Radkowsky Thorium Fuel Reactor concept.

15 231 Pa (Т ½ = 3х10 4 y): g/t Th SNF (PWR reactor) ppm 232 U (1 3 ppb in UO 2 SNF of LWR reactor): Content of 232 U >10 ppm remote handling is required.

16 Actinides relations in U nuclear cycle. 232 Th position is there.

17 Generation of actinides in the Th or U cycles. 232 Тh fertile (non-fissile), source of neutrons driver fuel ( 233 U, 235 U, 239 Pu) or ADS is required for start of operation. МА content in SNF may be high, especially in case of Pu rg. Another dangerous МА ( 231 Ра, Т½ = y) is formed.

18 Th-based SNF has more 79 Se (T½ =10 6 y), high burn up and heat generation ( 232 U, 90 Sr).

19 Radiotoxicity of FP in U or Th SNF in ALI (Ali = Annual limit intake). Red : more amount for Th cycle, blue more in U cycle.

20 Evidence of HLW disposal safety: Stability of U, Th, REEs ores and so on. Radioactive minerals study (pyrochlore, monazite, zirconolite ). Geochemical data for rocks around natural reactor in Gabon, Western Africa. Experiments on the waste forms stability, solubility, radionuclides sorption an so on. Theoretical calculations of performance assessment of HLW geological repository.

21 Different methods: age Ma U-Th system was closed for a long time. Crystalline matrices for L n An fraction isolation. 21

22 Only some long-lived RN: fission products: 79 Se, 99 Tc, 129 I, activated isotopes: 14 C, 36 Cl, and daughter isotopes of actinides decay can reach the biosphere.

23 Calculated doses from HLW repository

24 UO 2 - SNF, fresh and after interaction with aerated water. (Th,U,Pu,MA)O 2 : fluorite type structure. Very stable at SNF storage and disposal unlike UO 2 (UO 2 U 3 O 8 ). Th SNF: high content of Pa, Sr, Se, (long lived FP). Se: migration (SeO 4 ) 2-, reduction (Se 0,Se 2- ) deposition.

25 Solubility of ThO 2 in water. Role of complexes with CO 3 2- and colloid particles ( µm). Migration of Pa (IV), Pa (V) as colloids (Pa 5+ is close to Nb 5+ )

26 90 о C, water from J-13, filtration ~5 nm, 10-6 g / l: Sr, Tc, Cs, < U, < Th, <0.2 Np, Pu, Am. Contents of 90 Sr and 137 Cs decreased with time. Solubility of (Th,U)O 2 : 3x10-7 3x10-9 g/m 2 d. Value at least 100 times is lower than for UO 2. (Th,U)O 2 spent fuel, LWBR (Shippingport), 2 5% 235 UO MWt day / kg.

27 SNF reprocessing & HLW separation (TRUEX, DIAMEX, THOREX) allows: To reuse fissile actinides (U, Pu) To facilitate HLW safe isolation To enlarge a repository capacity due to heat & volume of waste decrease Main fractions: Cs-Sr, I, Tc, Np, MPG, Ln-An (Ln - REE Ce ; An - Am, Np, Cm)

28 PUREX: > t UO 2 SNF is reprocessed up to date. THOREX: ThO 2 is more stable than UO 2, (U,Pu)O 2. Additives of (0.05 M HF M AlNO 3 ) to HNO 3. In different schemes (production of U, U+Pu, U+Pu+Th) the different types of the waste forms are required.

29 Average composition of Ln-An fractions: 80-90% Ln % An and 60-70% Ln % Zr % An Ln: Nd>Ce>La~Pr,Sm; An: Am>Cm,Np. Special matrices for wastes are required. Zirconolite, Monazite which else? Ti,Sn,Zr-Pyrochlore & Fe-Garnet

30 Pyrochlore, [8] A [6] 2 B [4] 2 O [4] 6 O, Z=8, Fd3m, a=9-12å and Garnet, [8] A [6] 3 B [4] 2 T 3 O 12, Z=8, Ia3d, a = 11-13Å, structures consist of a network of BO 6 (pyrochlore) or BO 6 and TO 4 (garnet) linked by shared vertices. The voids are occupied by large cations with CN = 8. The AO 8 polyhedra shared edges with BO 6 and TO 4 : correlation between cationic sizes in the sites Natural pyrochlore Nb-Ta-Ti minerals containing up to 30 wt.% U, 9 wt.% Th. Minerals of the garnet group are silicates with low actinides content (< 0.1 wt.%).

31 View on pyrochlore- ( [8] A 2 [6] B 2 O 7, Fd3m) and garnet-type ( [8] A 3 [6] B 2 [4] T 3 O 12, Ia3d) structures. Structure Composition Interdependence.

32 Pyrochlore or garnet-based waste forms (matrices) for actinides (Th, U, Pu), minor actinides (Pa, Np, Am, Cm) and Ln-acinides. Contents of the waste elements can vary from 30 to 70 wt.%. Matrices can be produced by sintering or melting.

33 SAED of Ln-Snpyrochlore at Kr cation irradiation A = 0,12La + 0,24Ce + 0,1Pr + 0,38Nd + 0,08Sm +0,05Eu+0,03Gd

34 Effect of cationic radii ratio in [8] A and [6] B sites of pyrochlores on their critical T values at high energetic 1 MeV Kr 2+ irradiation.

35 HRTEM images of garnet before (a) and after (b d) irradiation with 1 MeV Kr to doses of (b) 0.09 dpa, (c) 0.14 dpa, (d) 0.22 dpa. Fe-garnets at 1 MeV Kr 2+ irradiation

36 Fe- and Al-garnets at Kr bombardment (2,3) and 244 Cm decay (1,4)

37 Corrosion resistance of ferrite garnets: Ca 1.5 GdCe 0.5 ZrFe 4 O 12, Ca 2.5 Ce 0.5 Zr 2 Fe 3 O 12, Ca 1.5 GdTh 0.5 ZrFe 4 O 12 water, 0.01M HCl, 0.01M NaOH, С, S/V=10m 1 Left: Th and Gd release from Ca 1.5 GdTh 0.5 ZrFe 4 O 12 and Cm from Ca 1.5 Gd 0.91 Cm 0.09 Th 0.5 ZrFe 4 O 12 (90 C). Right: SEM/BSE image of Ca 1.5 GdTh 0.5 ZrFe 4 O 12 garnet after its interaction with 0.01 M HCl solution for 30 days.

38 Good experience with U spent fuel. Pyrochemical reprocessing of Th-based SNF: hydrogenation of metal spent fuel chlorination of metal or oxide SNF evaporation of ThCl 4 and UCl 4 at heating electrolysis of molten salts precipitation of minor actinides and FP (Ln) fabrication of the fresh nuclear fuel waste isolation into glass, glass-ceramic, ceramic.

39 HLW (spent electrolite) from pyrochemical reprocessing P-Glass, unstable at high T. CEA, 2009 Phosphate precipitation of Ln and An Sodalite Vance et al, monazite Cl-apatite Vance et al., After 30 y of storage in clays ( 135 Cs) Low loading with HLW only 10 wt.%.

40 Research and commercial reactors with Th fuel, > 200 reactor - years (IAEA TECDOC-1450). Country Reactor Capacity Fuel composition Time Germany HTGR (Pebble bed) 15 MW(e) Th U, Coated Oxide & carbides Germany The same 300 MW(e) The same Germany BWR 60 MW(e) Fuel (Th,Pu)O 2 pellets UK, Sweden HTGR (Pin-in-Block Design) 20 MW(t) Th U Driver, Coated fuel particles, Oxide & Dicarbides USA HTGR 40 MW(e) The same USA (Prismatic Block) 330 MW(e) USA MSBR 7.5 MWt 233 U Molten Fluorides USA BWR (Pins) 24 MW(e) Th U Fuel Oxide USA LWBR PWR (Pins) 100 MW(e) Th U Driver Fuel, Oxide Pellets USA The same 285 MW(e) The same Canada MTR (Pins) MW Th U, Test Fuel India MTR Thermal 40 MW(t); 100 MW(t) Al U Driver Fuel, Th & ThO India PHWR (Pins) 220 MW(e) ThO 2 Pellets 1980 pres. India LMFBR (Pins) 40 MW(t) ThO 2 blanket 1985 pres.

41 Short-term (< 2020), reactors II III generations. (Russian approach. In: Internationalization of the NFC. 2008). Th - SNF of fast reactor contains ppm 232 U.

42 Medium term period (>2020). IV-generation reactors.

43 Characteristics of different reactors.

44 Very High Temperature Reactor (VHTR): The chief attraction of the VHTR concept is its ability to produce the higher temperatures (up to 1000 o C) needed for hydrogen production and some process heat applications. However, VHTRs would not permit use of a closed fuel cycle. Reference designs are for around 250 MW of electricity, or 600 MW of heat, with a helium coolant and a graphite-moderated thermal neutron spectrum. Fuel would be in the form of coated particles, formed either into blocks or pebbles according to the core design adopted. VHTR designs are based on prototype high-temperature gas-cooled reactors built in the United States and Germany, and much R&D has been completed. Challenges: developing improved temperature-resistant materials, fuel design + manufacture. No metal in active zone very high NF burn up.

45 Linear heat rate and burn-up of nuclear fuel for light water (LWR), fast (FR), and high temperature reactors (HTR); oxide fuel is the reference for LWR and HTR. In: NUCLEAR FUELS. CHAPTER 34. Konings, et al., Р

46 Examples of operated H T gas reactors.

47 ORNL/TM-2004/104. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR), May High fuel fabrication cost, but since 1960-s it is reduced > times to 6000$/kg (U-Pu MOX: $).

48 TRISO before and after irradiation Direct disposal of SNF with very high burn up. Fuel block with mass ~80 kg (including 4 kg of U or Th) contains 13 millions of TRISO particles.

49 In Th H-TR much more Ра, more Np, less Pu, Am, and Cm are formed comparatively with U cycle.

50 Advantage of usage HT and VHT gas reactors.

51 Molten Salt Reactor (MSR): In MSR, the fuel materials are dissolved in a circulating molten fluoride salt coolant. The liquid fuel avoids the need for fuel fabrication and allows continuous adjustment of the fuel mixture. The current concept is for a MW fast neutron reactor with a closed fuel cycle. This could be used for breeding with fertile thorium or for burning plutonium and other actinides. Molten salt chemistry, handling and corrosion resistance, as well as materials and the fuel cycle, are the most actual current R&D challenges. Fuel = coolant: LiF NaF BeF 2 ThF UF 4. Safety & simple management of high level waste.

52 Salt reprocessing scheme. Black / blue arrows salt route, orange arrows point out transfers through lead-bismuth bath and broad blue / violet arrows transfers through gas (SNETP, 2002). Fe-Ni-alloy Ceramic Different matrices for different waste stream

53 CONCLUSIONS: Approach to Management of Waste of Th cycle: 1) Open cycle. Waste is SNF itself (for LWR and VHTGR with very high burn up, > 100 GWt d / t). 2) Closed cycle with the SNF reprocessing by water-extraction (THOREX) or dry technologies. Main goal is extraction of rest or new formed fissile radionuclides ( 233,235 U; 239,241 Pu) for their usage for fuel fabrication (LWR, AHWR and FR with the lowest 232 U content). Waste forms are similar with the matrices developed for U cycle: glasses or glass-ceramics for fission products, ceramics for long-lived actinides, alloys for 99 Tc.

54 References. 1. Uranium-2007: Resources, production and demand. NEA-OECD. Report Management of recyclable fissile and fertile materials. NEA-OECD. Report Lung M.L., Gremm O. // Nuclear Engineering and Design Vol.180. P Kazimi M. // American Scientist Vol.91. N5. P Thorium fuel utilization: Options and trends. IAEA-TECDOC Jerden J.L., Cunnane J.C. Mat. Res. Soc. Symp. Proc Vol.757. II9-3.pdf. 7. Thorium fuel cycle Potential benefits and challenges. IAEA-TECDOC Internationalization of the nuclear fuel cycle: goals, strategies, and challenges. NAP GEN-IV International forum. Annual report The future of the nuclear fuel cycle. MIT Report Blue Ribbon Commission Updated Report. January Kopyrin A.A., Karelin A.I., Karelin V.V. Technology of production and radiochemical reprocessing of nuclear fuel. Atomenergoizdat (In Russian). 13. Laverov N.P., Yudintsev S.V., Livshits T.S. et al. // Geoch. Intern Vol.48. N1. P The Chemistry of the Actinide and Transactinide Elements. Eds.: Morss L.R., Edelstein N.M., Fuger J. Springer Vol. 6.