Fuel Cycle Options and Their Effects on Design and Performance of Saturated-Repository

Transcription

1 Fuel Cycle Options and Their Effects on Design and Performance of Saturated-Repository Advanced Summer School of Radioactive Waste Disposal with Socio-Scientific Literacy University of California, Berkeley August 3 7, 29 Nuclear Waste Management Organization of Japan (NUMO) Tomio Kawata Contents Nuclear renaissance and fuel cycle policies Is nuclear energy truly sustainable? Impact of fuel cycle option on HLW repository design and performance Conclusions 1 1

2 Nuclear renaissance and fuel cycle policies Dollar/b barrel Energy and the World Rapid energy demand growth in China and India, two world s most populous countries Peak oil problem Global warming Oil Price m 2 Million km Arctic Sea Ice NASA September Ice Extent 5 National Snow & Ice Data Center

3 Sustained use of nuclear energy is indispensable for our future generations Fossil En nergy Consumption A. E. Walter 1 12 kwh/y yr Pyramids First Ham mmurabi No Fossil Fuels Used Co onfucius Christ Mohammad Industrial Revolution No Fossil Fuels Left Year Past Present Future I Needs for Energy savings + Renewable /Alternative Energy + Nuclear Energy Wo orld Population, Nuclear Renaissance USA COL applications docketed for 26 reactors China 14 reactors under constr. 35 reactors planned Russia 37 new reactors in 22 roadmap 5 3

4 Light Water Reactor and Nuclear Fuel Nuclear Fuel Assembly Fuel Pellet (UO 2 ) Cladding Tube (Zircalloy) Nuclear Power Station (PWR) Fuel Rod 6 LWR fuel composition before and after utilization Fission U % U % Products 3-5% Pu ca. 1 % Residual U235 ca. 1 % Reusable Electricity Generation U238 > 9 % HLW in recycle option HLW in once- throughh option Fresh Fuel Spent Fuel 7 4

5 Fuel Cycle Policy Chosen by Countries with Large-scale Nuclear Power Program (>2 GWe) Country N. of Installed Capacity Fuel Cycle Policy NPP (GWe) Fast Reactor Program USA Direct disposal ABR development in GNEP (aborted?) France Closed cycle Japan Closed cycle Russia Closed cycle Germany Nucl. Phase-out - Korea (22) China (22) India (22) Phenix in operation GEN-IV prototype by 22 Monju prepairing for restart FaCT started (prototype by 225) BN6 in operation Construction of BN8 in progress Wait & See Design study in progress (KALIMER) Closed cycle CEFR under constr. (29) CPFR by 22 Closed cycle PFBR under constr.(21) 4 Commercial FBRs by 22 Installed capacity as of September 28 8 World Installed Capacity by Fuel Cycle Policy % 47% DD countries other than USA CC Countries 27% USA DD countries other than USA 19% 25% USA 56% CC Countries Direct Disposal (DD) Closed Cycle (CC) 9 5

6 Chopped fuel & hull FP U Pu Reprocessing plant Uranium purification Uranium Conversion (de-nitration) Spent fuel receipt Cask Fuel chopping machine Chemical separation UO 3 powder storage Storage pool Dissolver MOX powder storage Spent fuel assembly Hull Vitrification High-level liquid waste Plutonium purification U-Pu Co-conversion 1 Reprocessing plants in the world B25 plant 15 t/y THORP 9 t/y (UK) RT-1 (Russia) 4 t/y Pilot plant (China) 25 t/y World capacity = 5,5 t-u/y 8 t/y Rokkasho Tokai (Japan) UP2-8 UP3 (France) total 17 t/y Trombay 6 t/y PREFRE-1 1 t/y PREFRE-2 1 t/y (India) in operation under planning 11 6

7 Reprocessing plants in the World UP2-8 and UP3 in La Hague (France) THORP in Sellafield (UK) Tarapur Repr. Plant Kalpakkam Repr. Plant (PREFRE-1) (PREFRE-2 or KARP) (India) Rokkasho Reprocessing plant (Japan) 12 Enhanced proliferation resistance in Japanese reprocessing plants European Reprocessing Plants UP2-8, UP3 THORP Spent fuel Reprocessing plant & conversion facility U 3 O 8 or UO 3 PuO 2 Japanese Reprocessing Plants Tokai RP Rokkasho RP Spent fuel Reprocessing plant & conversion facility UO 3 MOX (Pu:U=1:1) 13 7

8 Spent fuel Plutonium and MOX fuel eprocessing R Uranium Plutonium nitrate solution MOX powder U235 MOX pellets Pu.7% 3-5% 4-9% U % UO 3 or U 3 O 8 Natural uranium LWR LEU fuel LWR MOX fuel FBR MOX fuel 14 Japan unique case Only one country with full-scope nuclear cycle program in Non-Nuclear Weapon States under NPT First country with nuclear power and fuel cycle program to qualify for Integrated Safeguards I am pleased to note that Japan has become the first State with an advanced nuclear cycle to qualify for integrated safeguards Statement by IAEA DG El Baradei to 24 IAEA General Conference (2 September 24) LWRs Enrichment Reprocessing MOX Fuel FBR 15 8

9 Is nuclear energy truly sustainable? 16 Sustainability Sustainability of fuel supply Sustained availability of waste (HLW) repositories + Economical competitiveness Safety Compatibility with nonproliferation norms Waste repositories are very precious social assets! 17 9

10 Calculated lives for various energy sources 192 years 2,5-5, years 67 years 85 years 42 years Coal Oil LNG Uranium LWR Fossil Fuel Oncethrough Uranium Multi-recycling in FBR Calculated life = Identified resources/current annual consumption Uranium 23: Resources, production and Demand 18 Three options to extract inexhaustible energy from fission LWR once-through + uranium recovery from sea-water Uranium inventory in sea-water = 4.5 billion tons FBR + multi-recycling of U-Pu-MA Current world stock of uranium (1.5 million tons of DU + RU) fuels 5 reactors with 1 GWe capacity for 3, years (1 GWa from 1 ton U) Thorium Cycle Thorium resource = 5 x uranium resource 19 1

11 Fuel Cycle Options in U-Pu System LWR cycle Once-through (direct disposal) Closed cycle (reprocessing & recycling) Complete closed cycle with FBR Reprocessing & Pu multi-recycling Reprocessing & Pu+MA multi-recycling 2 Material balances in two LWR cycle options Uranium Ore Mining & Milling Waste 1, t Natural Uranium 17 t Depleted Enrichment Uranium 15 t Enriched Uranium 2 t Electricity 1, MWa LWR Spent Fuel 2 t LWR once-through Waste Uranium Ore Mining & Milling Waste 85, t Natural Uranium 15 t Depleted Enrichment Uranium 13 t EU 17 t Electricity 1, MWa Waste Spent Fuel 2 t LWR Uranium saving = 15 % Recovered U 16 t (storage) MOX 3 t Reprocessing & MOX fab. Vitrified HLW 6 8 t (FP 1t) LWR reprocessing & recycling TRU waste 21 11

12 Material balance in FBR cycle Electricity 1, MWa Natural, depleted, or recovered uranium 1t 1 t Plutonium FBR Spent Fuel 1-2 t FBR equilibrium cycle Waste 1 GWa electricity Uranium & from 1 ton U Plutonium Current world 1-2 t stock of DU+RU is around 1.5 million tons which is enough to fuel 5 reactors with 1 GWe capacity Reprocessing & Fuel Fab. (Recycle) Vitrified HLW 6 8 t (FP 1 t) TRU Waste for 3, years Either uranium mining or uranium enrichment is no longer necessary 22 Demerits of continued reliance on LWR system Extremely low uranium utilization efficiency (1% at most) Endless accumulation of enrichment tail Plutonium mine problem in direct disposal Larger space requirement for HLW repository in direct disposal Accumulation of higher isotopes of Pu and MA in reprocessing & recycling Endless reliance on LWR system may not be desirable from back-end aspects! Merit of FBR cycle with MA recovery/burning High uranium utilization efficiency (> 6%) Eliminate i necessity of uranium mining i and enrichment Burn legacy waste of LWR era (DU, Pu and MA) Higher thermal efficiency (FBR 4% vs. LWR 33%) Smaller space requirement for HLW repository Demerits of FBR cycle: Economical penalty Proliferation risk 23 12

13 Cylinder yard at Paducah 64, cylinders in total are stored at Paducah and Portsmouth sites (nearly 7, tons of DUF 6 ) Uranium processing plants have been constructed at two sites to convert unstable DUF 6 into stable DU 3 O 8 However, it takes 25 years to complete whole processing No plan to deal with DU 3 O 8 (prolonged storage?) 24 Problem of direct disposal from nonproliferation aspects 1 years later, access becomes easier and plutonium becomes more attractive for weapon use Safeguard is required eternally for direct disposal repositories Exposure in 8 hours at 1 m away fr rom the side of waste package 1 msv Cooling time, Years Radiation exposure ue to WG-Pu Relative val Neutron emission Heat generation Cooling time, Years Plutonium properties 25 13

14 Impact of fuel cycle options on HLW repository design and performance (saturated-repository) 26 Examples of Reprocessing Options Spent Fuel Separation mode U Pu FP + MA Type of HLW MA bearing HLW Spent Fuel Excess U U + Pu (+ Np) FP + MA MA bearing HLW Spent Fuel Spent Fuel Excess U U + Pu (+ Np) MA FP MA Cs + Sr FP other than Cs/Sr MA free HLW Low heat HLW 27 14

15 Major Variables for Repository Design Waste Form Spent Fuel, Vitrified HLW Natural Barrier Host Rock Crystalline rock, Sedimentary rock, Salt dome etc. Repository Location Saturated zone, Unsaturated zone Type of Ground-water Freshwater, Saline water Engineered Barrier Canister, Overpack, Buffer Canister, Drip shield, Invert Waste Emplacement Vertical, Horizontal 28 HLW disposal in saturated-repository (Multi-barrier concept) Host rock Deep underground Buffer location (deeper than 3 m) Plastic/swelling Isolation from surface perturbations Low water flow rate Reducing environment Natural Barrier (bentonite + quartz sand) Plastic/swelling mechanical barrier Low permeability hydraulic barrier Radionuclide sorption Colloid filter Overpack (carbon steel) Containment for more than 1, years Vitrified HLW canister Extremely low solubility Engineered Barrier System (EBS) 29 15

16 Major design parameters to govern waste loading density in the repository Physical dimension of waste package vitrified HLW or spent fuel Mechanical strength of host rock hard rock or soft rock Decay heat vitrified HLW or spent fuel UO 2 or MOX cooling time before burial reprocessing method In general, decay heat is dominating design parameter to govern waste loading density 3 Reprocessing vs. Direct disposal equivalent Vitrified HLW canisters with overpack Spent fuel canister (NAGRA) 31 16

17 Major Heat Sources and Half Lives Fission Products Sr-9 (28.8 y) Cs-137 (3 y) Actinides Pu-238 (87.7 y) Pu-24 (656 y) Am-241 (432 y) Pu-241 (14.4 y) Cm 244 (18.1 y) 32 Comparison of Decay Heat - Vitrified HLW vs. Spent Fuel - Decay Heat, W/t-U Vitrified HLW GWd/t 12 Cm Reprocessing Am 1 after cooling for 4 years Pu 8 FP Time after reprocessing, year Decay Heat, W/t-U LWR UO 2 Spent Fuel GWd/t Cm Am Pu FP Time after discharge, year 33 17

18 HLW Disposal Concept in Japan Bentnite Buffer Max. Buffer Temp. < 1 degree C Waste Canister Overpack Host Rock 34 Variation of max. temperature in buffer material for direct disposal with varied allocated space Case1 - Allocated space same as the reference case of vitrified HLW disposal Case2 - Allocated space = Case1 x 2 Case3 - Allocated space = Case1 x 4 Host Rock a Waste Package Cupper Overpack Buffer b [deg. C] Temperature [ Time after burial [y] a b Case1 Case2 Case

19 Peak temperature of buffer material calculated for direct disposal 18 Sedimentary rock 16 Horizontal emplacement Buffer temperature limit Allocated space (relative value to the reference disposal space for vitrified HLW) Peak Te emperature, deg g.c 36 Impact of fuel cycle options on repository loading - Reprocessing vs. Direct disposal - Study JAEC study (Japan) JAEA study (Japan) ANDRA Dossier 25 (France) ONDRAF SAFIR-2 (Belgium) ANL/AFCI (USA) Improvement factor for waste loading density (a) (b) (c) (a) Constraints from other factors than thermal design are involved (b) Disposal of MOX or HLW from MOX reprocessing is included (c) Case for 99.9 % removal of U, Pu, Am and Cm 37 19

20 Case to continue reprocessing beyond 21 French Study Space for vitrified waste to be generated in 21 and beyond Case to terminate reprocessing in 21 Space for spent fuel disposal to begin in 21 Space for vitrified waste generated before 21 D. Greneche, AREVA 38 Footprints of waste repositories - Direct disposal vs. Reprocessing & recycling - 16 Report of Long-Term Plan Deliberation Committee, Nov. 24, JAEC 1km x HLW Repository (geologic disposal) m 3 Waste volume, 1 4 m Direct disposal x 9 4 HLW 2 HLW 1/2 HLW (Note) Low-level wastes from decommissioning of reprocessing and MOX plants are included Surface pit disposal Disposal in intermediate depth Geologic disposal Repository (geologic disposal) Reproc. & recycle 39 2

21 Typical repository layout Disposal panel for HLW (HLW/TRU co-location option) Disposal panel for TRU waste Hard rock Pit disposal Tunnel disposal Access tunnel HLW TRU waste Soft rock Emplacement of waste package 4 Japan France 41 21

22 Reprocessing vs. direct disposal Compared with reprocessing case, direct disposal requires several times larger repository space because of the additional heat from Pu and Am In reprocessing case, secondary low-level waste in relatively large volume is generated However, disposal space needed for lowlevel waste is far smaller than that for HLW In France, they succeeded in significant volume reduction for secondary waste 42 HLW from MOX fuel reprocessing in LWR and FBR cycles 43 22

23 Isotopic composition of various plutonium Is sotopic composit tion 1% 8% 6% 4% 2% % Weapon -grade Pu Pu from LWR spent UO2 (*) Pu from LWR spent MOX (*) Pu in FBR Equilibrium cycle (**) Pu242 Pu241 Pu24 Pu239 Pu238 (*) Burnup 33 MWd/t (**) Core and blanlet blend 44 Decay heat of vitrified HLW from spent MOX fuel t, kw/gwy Decay hea Significant heat increase due to Am and Cm Heat from Am becomes overwhelming in delayed reprocessing LWR-MOX Early Reprocessing 45 GWd/t Reprocessing after cooling for 4 years Cm Am FP Years after reprocessing Decay hea at, kw/gwy Decay heat, kw/gwy LWR-UO 2 48 GWd/t Reprocessing after cooling for 4 years Cm Am FP Years after reprocessing LWR-MOX Delayed Reprocessing 5 45 GWd/t Reprocessing after cooling for 3 years Cm Am FP Years after reprocessing 45 23

24 Decay heat of various HWL mixtures Disposal of HLW from MOX by itself would be problematic because of its large decay heat Mixing with HLW from UO 2 will mitigate the problem Decay heat, W/t-HM UO 2 (early reprocessing) UO % MOX (early repr.) UO % MOX (delayed repr.) Time after reprocessing, year 46 Decay heat of vitrified HLW in FBR cycle Decay heat at 5-year cooling in FBR cycle with MA recovery is smaller than in LWR cycle Decay heat, kw/gwy LWR-UO 2 48 GWd/t Reprocessing after cooling for 4 years Cm Am FP Years after reprocessing t, kw/gwy Decay heat Commercial FBR 15 GWd/t Without MA Recovery GWd/t Cm Reprocessing 35 Am after cooling 3 FP for 4 years Years after reprocessing, kw/gwy Decay heat Commercial FBR 15 GWd/t 9 % MA Recovery GWd/t Cm 35 Reprocessing Am after cooling 3 FP for 4 years Years after reprocessing 47 24

25 Fission yields of U235 and Pu239 ) ission yields (% F Zr Tc Sr Pu239 I Cs U Mass 48 Number of glass canisters/g GWy Number of glass canisters per unit power generation LWR-UO 2 48GWd/t LWR-MOX 45GWd/t Reprocessing after cooling for 4 years Max. centerline temperature < 5 deg.c Heat loading < 2.3 kw for cooling by natural convection Heat generation from MA Heat generation from FP FBR-mix 15GWd/t(core) 2GWd/t (ax. blkt) 6.4GWD/t (rad. blkt) 49 25

26 Peak Tempe erature of Buffer, de eg.c Effect of increase in decay heat of vitrified HLW Decay heat Reference case x 2 Reference case x 1.5 Temperature limit: 1 deg.c 2 Sedimentary rock Horizontal emplacement Allocated space (relative value to the reference case) 5 Comparison of repository space per unit power production in various fuel cycle options 1.LWR once-through 2.LWR-UO2 reprocessing 3.LWR-UO2 reprocessing + MA removal 4.LWR-MOX reprocessing 5.LWR-MOX reprocessing + MA removal 6.FBR recycling 7.FBR recycling + MA removal Repository space per unit power production (m 2 /GWy) Source: JAEA 51 26

27 HLW from Pu-bearing fuel (LWR-MOX and FBR fuel) Relatively l large heat generation from MA Delayed reprocessing results in the accumulation of Am241 which hardly decays during the storage period (half life is 432 years!) 9% removal of MA significantly reduces heat load of HLW In future FBR cycle, introduction of MA removal and burning capability would improve the waste loading density by a factor of 2 in comparison with current LWR cycle 52 MA removal for reduction of radiotoxicity 53 27

28 Radiotox xicity (relative va alue) Reduction of Radiotoxicity Radiotoxicity of Uranium Ore LWR spent fuel Vitrified HLW (99.9% MA removal) Vitrified HLW (without MA removal) Cooling Time (year) 54 Ra adiotoxicity, TBq/GWy Relative radiotoxicity shown in linear scale 1, 5 9, LWR once-through LWR reprocessing 8, FBR MA recycle 4 7, 6, 5, 4, 3, 2, 1, Ra adiotoxicity, TBq/GWy , Cooling time, year 1, 2, 3, 4, 5, Cooling time, year 55 28

29 Annual individ dual dose μsv/y y Dominant radionuclides in safety assessment 1, micro Sv/y SAFIR2 (SF) SAFIR2 (HLW) D25 (SF) D25 (HLW) SR97 H12 (HLW) 1, 1, 1, 1,, Time after repository closure, year EN22 (SF) EN22 (HLW) SR97 SAFIR2 (SF) SAFIR2 (HLW) EN22 (SF) EN22 (HLW) H12 (HLW) I-129 Se-79 Se-79 I-129 Th-229 Se-79 Sn-126 C-14 Cl-36 I-129 Th-229 Pd-17 I-129 I-129 Se-79 Cs-135 Th Effect of MA removal on radiotoxicity and dose 99.9% removal of MA significantly reduces radiotoxicity ( reduction of potential risk) However, radiotoxicity is not regarded as a safety index to be used in the safety assessment for licensing Usually, dose evaluation to the critical group is required in the safety assessment Because of their low solubility and low mobility, MA (usually in the form of oxide) are not dominant nuclides for dose evaluation 57 29

30 Conclusions Direct disposal option requires several times larger space than recycle option For Pu-bearing fuel, conventional reprocessing will results in the production of HLW with relatively large heat generation due to Am241 accumulation, and a special attention is needed to deal with this fact In order to remove excess heat load from MA, the removal up to 9 % is sufficient, whereas the removal by 99.9 % is required for radiotoxicity reduction In the future FBR cycle, 9 % MA removal and subsequent burning allow for a significant improvement in the repository utilization efficiency for the same electricity generation In practical view, MA removal for heat reduction is more important than that for radiotoxicity reduction 58 Fast (Breeder) Reactor Program is reviving! Japan - Monju preparing for restart -FaCTprogram in progress (prototype by 225) Russia - BN6 in operation - BN8 under constr. India - PFBR under constr. - 4 CFBRs by 22 France - Phenix in operation - Gen-IV prototype by 22 China - CEFR constr. completing - CPFR by 22 USA - ABR study??? Korea - Design study in progress (KALMER) 59 3

31 Key factors for future FBR cycle (other than safety and economy) Multiple recycling MA removal and burning Co-recovery (no pure Pu) Sustainability of fuel supply Sustainable availability of HLW repositories Enhancement of proliferation resistance 6 Who e er aspiring struggles on, For him there is salvation Goethe s Faust 61 31