Thorium Fuel Cycles & Heavy Water Reactors AECL Experience

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1 Thorium Fuel Cycles & Heavy Water Reactors AECL Experience Energy From Thorium Event CNS UOIT B. P. Bromley Advanced Reactor Systems Computational Reactor Physics AECL - Chalk River Laboratories March 22, 2013 UNRESTRICTED / ILLIMITÉ

2 Opening Remarks There s nothing magical or mysterious about thorium except: 3 times abundant as uranium in the earth s crust a large resource. U-233 (bred from Th-232) has a high 2.2, in both thermal and fast neutron energy spectrum; can be used for a breeder reactor. Pu, Am, Cm, etc. production with Th-based fuels will much lower. Any reactor (fast or thermal) can be adapted to use thorium. Thermal reactors can operate with lower fissile wt%. For thermal spectrum reactor, we want: Minimal parasitic neutron absorption; maximum neutron economy. Maximum burnup for Th-based fuel for a given fissile content. OTT (Once Thru Thorium) Cycle SSET (Self-sustaining Equilibrium Thorium) Cycle Topping fuel cycles: Th (new + recycled) + (U/Pu) (new + recycled) UNRESTRICTED / ILLIMITÉ 2

3 Fundamental Advantage of Heavy Water Heavy water has the highest moderating ratio ( s / a ). Slows down neutrons with minimal absorption. Better than H (in H 2 O), better than C (in graphite). Can maximize neutron economy, in a thermal-spectrum reactor. Save neutrons for fission and breeding new fuel. HWR can run on natural uranium and achieve good burnup. ~7,500 MWd/t in a CANDU PT-HWR UNRESTRICTED / ILLIMITÉ 3

Pressure Tube Heavy Water Reactors (PT-HWR) - Advantages Pathway Canada Chose AECL Pursued. Excellent neutron economy. High conversion ratios (C.R.>0.8).

Higher fuel burnup for a given enrichment. 30% more burnup than 3-batch refuelling. Maximize uranium utilization (kwh/kg-u-mined). High capacity factors (0.8 to 0.95).

4 Pressure Tube Heavy Water Reactors (PT-HWR) - Advantages Pathway Canada Chose AECL Pursued. Excellent neutron economy. High conversion ratios (C.R.>0.8). Can operate on natural uranium (NU). High fuel utilization; conservation of resources. Continuous On-line refuelling. Low excess reactivity. Higher fuel burnup for a given enrichment. 30% more burnup than 3-batch refuelling. Maximize uranium utilization (kwh/kg-u-mined). High capacity factors (0.8 to 0.95). Flexibility in fuel loading one or more fuel types can be used. Modular construction. Short, relatively simple fuel bundle design. Pressure tubes; replaceable; reactor can be refurbished. Local fabrication (do not need heavy forgings). UNRESTRICTED / ILLIMITÉ 4

5 PT-HWR Operational Technology. Future HWR variants. Potential for further improvements. Use R&D to find them. UNRESTRICTED / ILLIMITÉ 5

6 PT-HWR / CANDU Reactors Designed to maximizes neutron economy. Flexible in use of fuel types. An existing, proven, and operational technology. Supply chain in place. Design naturally lends itself to implementation of Th-based fuels. Thorium-based fuels have been tested in PT-HWR (NPD-2). Thorium bundles have been used in India (in their PT-HWRs). Power flattening for start-up cores; alternative to DU. Practical implementation time should be relatively short. AECL / CRL has helped develop and prove this technology, and is continually exploring technology improvements to facilitate implementation and expansion of thorium-based fuel cycles. Emphasis on use of solid fuel forms. UNRESTRICTED / ILLIMITÉ 6

Overview of AECL Experience AECL has over 50 years of extensive experience with Thoria-based fuels Investments made in thorium fuel cycle R&D since the late 1950 s First irradiation conducted in 1962

7 Overview of AECL Experience AECL has over 50 years of extensive experience with Thoria-based fuels Investments made in thorium fuel cycle R&D since the late 1950 s First irradiation conducted in 1962 and the most recent in 2005 Experience includes Fuel Fabrication. Irradiation testing. Post Irradiation Examination. Thorium fuel reprocessing. Waste management. Critical experiments. Reactor physics. Conceptual design studies. Economic analyses. System studies. UNRESTRICTED / ILLIMITÉ 7

Recycle With U-233 Recycle U-233 + Pu Once-through LEU/Th

8 Innovation Thorium in CANDU / PT-HWR Evolution PT-HWR Canadian SCWR Build U-233 resource U Pu Once-through Pu/Th With U-233 Recycle With U-233 Recycle U Pu Once-through LEU/Th Once-through Pu/Th Years AECL - OFFICIAL USE ONLY / À USAGE EXCLUSIF - EACL 8

9 Decay heat (GW) Long-term Impact Decay heat in spent fuel is a main parameter in determining the capacity of a long term disposal facility Current global cycle, LWRs + HWRs Transition to once-through thorium in CANDU Transition to fast reactors Transition to Th with U-233 recycle in CANDU Once-through thorium gives the same reduction as fast reactors Once-through thorium gives 50% reduction over current cycle Years since end of scenario Thorium with U-233 gives a 75% reduction over the current cycle AECL - OFFICIAL USE ONLY / À USAGE EXCLUSIF - EACL 9

10 Fabrication Generally, AECL has targeted a solid solution of Thoria and the fissile additive. Many techniques are capable of achieving this and they fall into two main categories: 1. Solution blending sol gel, co-precipitation Mixing at the atomic level 2. Mechanical mixing co-milling, high-intensity mixing Often not a perfect solid solution Must achieve mixing on the scale of individual particles UNRESTRICTED / ILLIMITÉ 10

11 Thorium Pellet Structure Granular Homogeneous UNRESTRICTED / ILLIMITÉ 11

12 Thoria Irradiation Experience at AECL Thoria irradiations ongoing since early 1960s Irradiations in NRX, NRU and WR-1 research reactors. Irradiations in NPD-2 ~20 MW e prototype PT-HWR. Pure ThO 2, (U,Th)O 2, and (Pu,Th)O 2 NRU still operational. Irradiation Facility # of experiments Irradiation Time frame NPD NRX NRU WR UNRESTRICTED / ILLIMITÉ 12

13 NRX, NRU, WR-1, NPD NPD-2 WP-1 NRU NRX UNRESTRICTED / ILLIMITÉ 13

14 Critical Experiments AECL: long history of critical experiments involving thorium-based fuels. Three sets of experiments, dating back to the 1960 s HEU/Th ( ) Pu/Th (1986) U-233/Th (1990s) Performed in the ZED-2 (Zero Energy Deuterium) critical facility at Chalk River Laboratories. Reaction rate / foil data. Reactivity changes due to X X = coolant density, temperature, etc. Verifies physics; validate computer codes. UNRESTRICTED / ILLIMITÉ 14

$4 Checkerboard Seed- Blanket Cores Row\Col 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 Row\Col A 0 0 0 0 0 0 0 0 B B B B B B 0 0 0 0 0 0 0 0 A B 0 0 0 0 0 B B B S S S S S S B B B 0 0 0 0$

15 Alternative Fuel Bundle and Core Design Options Fissile Utilization, Relative to Natural Uranium CANDU Heterogeneous, Mixed Bundle Homogeneous Bundle NU Checkerboard Seed- Blanket Cores Row\Col Row\Col A B B B B B B A B B B B S S S S S S B B B B C B S S B S S B B S S B S S B C Annular Seed-Blanket Cores Row\Col Row\Col A B B B B B B A B B B B S S S S S S B B B B C B S S S S S S S S S S S S B C Checkerboard Core Designs D B S S B B S S B B S S B B S S B D E 0 0 B S S B S S B B S S B B S S B S S B 0 0 E F 0 0 B S B B S S B B S S B B S S B B S B 0 0 F D B S S S S S S S S S S S S S S B D E 0 0 B S S S S S S S S S S S S S S S S B 0 0 E F 0 0 B S S S S S S S S S S S S S S S S B 0 0 F Annular Core Designs G 0 B S B S S B B S S B B S S B B S S B S B 0 G G 0 B S S S S S S S S S S S S S S S S S S B 0 G H 0 B B B S S B B S S B B S S B B S S B B B 0 H H 0 B S S S S S S S S S S S S S S S S S S B 0 H J B S S S B B S S B B S S B B S S B B S S S B J J B S S S S S S S S S S S S S S S S S S S S B J K B S S S B B S S B B S S B B S S B B S S S B K K B S S S S S S S S S S S S S S S S S S S S B K L B S B B S S B B S S S S S S B B S S B B S B L L B S S S S S S S S S S S S S S S S S S S S B L M B S B B S S B B S S S S S S B B S S B B S B M N B S S S B B S S B B S S B B S S B B S S S B N O B S S S B B S S B B S S B B S S B B S S S B O M B S S S S S S S S S S S S S S S S S S S S B M N B S S S S S S S S S S S S S S S S S S S S B N O B S S S S S S S S S S S S S S S S S S S S B O Hafnium Tube 3 mm thick 3.8% Pu 96.2% Th P 0 B B B S S B B S S B B S S B B S S B B B 0 P P 0 B S S S S S S S S S S S S S S S S S S B 0 P Q 0 B S B S S B B S S B B S S B B S S B S B 0 Q R 0 0 B S B B S S B B S S B B S S B B S B 0 0 R S 0 0 B S S B S S B B S S B B S S B S S B 0 0 S T B S S B B S S B B S S B B S S B T Q 0 B S S S S S S S S S S S S S S S S S S B 0 Q R 0 0 B S S S S S S S S S S S S S S S S B 0 0 R S 0 0 B S S S S S S S S S S S S S S S S B 0 0 S T B S S S S S S S S S S S S S S B T Zr Rod PT CT U B S S B S S B B S S B S S B U U B S S S S S S S S S S S S B U V B B B S S S S S S B B B V V B B B S S S S S S B B B V W B B B B B B W Row\Col Row\Col W B B B B B B W Row\Col Row\Col Moderator AECL - OFFICIAL USE ONLY / À USAGE EXCLUSIF - EACL 15

16 Fissile Utilization Relative to PT-HWR-NU Potential to Increase Utilization Achieve ~ 20% to 100% higher utilization of fissile fuel than PT-HWR with NU fuel in an OTT cycle PT-HWR NU LEU/Th - 8-Th 35-LEU/Th LEU/Th - ZrO2 Rod 21-LEU/Th LEU/Th - ZrO2 Rod 35-Pu/Th-8-Th Pu/Th 35-Pu/Th - ZrO2 Rod Pu/Th 21-Pu/Th - ZrO2 Rod Volume Fraction of Initial Fissile in Bundle IHM (LEUO2, PuO2, ThO2) UNRESTRICTED / ILLIMITÉ 16

Thorium in China Evolutionary Approach with HWRs Uranium resources limited. Use of Canadian know-how - PT-HWRs. Use NUE in CANDU-6 (2012-2014). RU~0.

Thorium-based fuels in EC6 ( 2020). Collaborate/cooperate w/ Canada. Simple, evolutionary design first, based on 43-element bundle carrier.

17 Thorium in China Evolutionary Approach with HWRs Uranium resources limited. Use of Canadian know-how - PT-HWRs. Use NUE in CANDU-6 ( ). RU~0.9 wt%; DU~ 0.25 wt% U-235/U NUE ~ 70% RU + 30% DU. Behaves the ~same as NU in CANDU. Use RU in dedicated EC6 (by ~2019). Thorium-based fuels in EC6 ( 2020). Collaborate/cooperate w/ Canada. Simple, evolutionary design first, based on 43-element bundle carrier. LEU in smaller outer 35 pins. Th in larger inner 8 pins. Core could be mix of NU, RU and Th-based bundles. Build up inventory of U-233 in spent fuel. Recycle in future. UNRESTRICTED / ILLIMITÉ 17

18 Summary Thorium has a great potential benefit for sustainability, safety, and waste management. Goals can be achieved by current commercial reactor designs. We don t need to wait, at least not long. PT-HWR s are operational today, and can be adapted for thorium. Small design changes can be implemented quickly. More R&D to enable more substantial design changes. R&D that will enable practical engineering solutions. AECL has 50 years experience in thorium fuel cycles: Reactor and fuel design; alternative concepts. Fuel fabrication. Irradiations + Critical Experiments. Reprocessing / Recycling, Waste management. Development, Testing & Validation of Analysis Tools. Economics and system analyses. UNRESTRICTED / ILLIMITÉ 18

19 More Info Visit: UNRESTRICTED / ILLIMITÉ 19

20 Acknowledgements Bronwyn Hyland Jeremy Pencer Holly Hamilton Laurence Leung CRL Library and Report Centres Various staff in Fuel Development Branch Computational Reactor Physics Branch. Applied Physics Branch (ZED-2 Facility) Thermal-hydraulics Branch Fuel Channel and Fuel Channel Safety Branch UNRESTRICTED / ILLIMITÉ 20

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22 Fast Fission in Fertile Isotopes U-238, Th-232 U-238 Th-232 UNRESTRICTED / ILLIMITÉ 22

23 AECL Work on ADS / HFFR for U-233 Production from Th Design studies, watching briefs, economic assessments. Accelerator-Drive Systems (ADS) Spallation fast neutron source driving U or Th blanket. Hybrid Fusion Fission Reactors (HFFRs) 14-MeV D-T fusion neutrons driving U, U/Th and Th blankets. Alternative to reactor-based breeders; high support ratio (10:1). Complement existing fleet of high converter PT-HWRs UNRESTRICTED / ILLIMITÉ 23

24 Neutron Production by Spallation 1 GeV protons or deuterons on Pb/Bi or U target ~20 neutrons per proton (Pb), ~ 40 neutrons per proton (U) UNRESTRICTED / ILLIMITÉ 24

25 Neutron Production in Fissile Isotopes Variation of neutron production per neutron absorted. Isotope Thermal Spectrum** Fast Spectrum*** U U Pu Pu Spectrum-Averaged Neutron Production ( ) for Various Fissile Isotopes ** Approximate range of values in a thermal-spectrum reactor. *** Approximate range of values in fast-spectrum reactor. UNRESTRICTED / ILLIMITÉ 25

26 Overview AECL has over 50 years of extensive experience with Thoria-based fuels Investments made in Thoria fuel cycle R&D since the late 1950 s First irradiation conducted in 1962 and the most recent in 2005 Experience includes Irradiation in Nuclear Power Demonstration reactor (NPD) and 3 experimental reactors Manufacturing fuels with a wide range of compositions and pellet geometries using both novel and traditional fabrication techniques Post Irradiation Examination (PIE) studies Knowledge gained from various experiments has been fed into new experiments Results indicate that Thoria fuel has always performed comparably with UO 2 and in some cases demonstrated superior performance UNRESTRICTED / ILLIMITÉ 26

27 Fabrication Fissile Additive Generally, AECL has targeted a solid solution of Thoria and the fissile additive Many techniques are capable of achieving this and they fall into two main categories: 1. Solution blending sol gel, co-precipitation Mixing at the atomic level 2. Mechanical mixing co-milling, high-intensity mixing Often not a perfect solid solution Must achieve mixing on the scale of individual particles UNRESTRICTED / ILLIMITÉ 27

28 Fabrication Sol-Gel Microspheres from 15 to > 1000 μm UNRESTRICTED / ILLIMITÉ 28

29 Experiment Summary Table Irradiation Facility # of experiments Irradiation Time frame NPD NRX NRU WR Each experiment consisted of a series of irradiations Other irradiations were done and reported in the literature UNRESTRICTED / ILLIMITÉ 29

30 Granular Pellet Structure 95% Dense UNRESTRICTED / ILLIMITÉ 30

31 Thoria Irradiation Experience at AECL Thoria irradiations have been ongoing since early 1960s Irradiations in NRX, NRU and WR-1 research reactors as well as NPD a 20 MW e power reactor Pure Thoria, Thoria-UO 2 & Thoria-PuO 2 UNRESTRICTED / ILLIMITÉ 31

32 Post Irradiation Examination PIE conducted for most of the individual irradiation Thoria fuels Typical PIE data depends on the nature of testing and program objectives and can include: Visual exam Element profilometry Axial gamma scanning Element gas puncture and fission gas analysis Burnup analysis Sheath metallographic exam Fuel pellet ceramographic exam α -ß-γ autoradiography UNRESTRICTED / ILLIMITÉ 32

33 Six Thoria-PuO 2 Bundles (1) Irradiation performed in NRU 36 element Bruce type fuel bundle wt% Th and 1.53 wt% Pu in (Th, Pu)O 2 Objectives To verify the ability of (Th, Pu)O 2 fuel to operate at significant power outputs to burnups of 42 MWd/kgHE To examine the power-ramp performance of Zircaloy clad (Th, Pu)O 2 fuel with ES-242 and siloxane sheath (CANLUB) coatings To determine fission-gas release To examine micro-structural changes in the fuel UNRESTRICTED / ILLIMITÉ 33

Six Thoria-PuO 2 Bundles (2) Maximum sustained powers from 49-75 kw/m Burnups to 45 MWd/kgHE (also maximum power bundle) Fission products accumulate in fuel grain

34 Six Thoria-PuO 2 Bundles (2) Maximum sustained powers from kw/m Burnups to 45 MWd/kgHE (also maximum power bundle) Fission products accumulate in fuel grain boundary which limit fuel performance Low gas release Low sheath strain Significant % of PuO 2 present as agglomerates Outer element in Bundle ADC-1 UNRESTRICTED / ILLIMITÉ 34

35 Thoria Demonstration Irradiation PIE Higher than expected gas release due to granular structure of pellets (WR tests) Granules UNRESTRICTED / ILLIMITÉ 35

36 High-Density, Homogeneous Thoria 1.5 % U-235, 35 MWd/kgHE 48 kw/m Max Low gas release Low sheath strain Irradiation is ongoing UNRESTRICTED / ILLIMITÉ 36

37 Conclusions AECL has over forty-seven years of experience with Thoria-based fuel irradiations, with burnups up to 47 MWd/kgHE AECL has extensive experience with Thoria fuels having a wide range of fuel compositions and pellet geometries Successful fabrication technology has been developed and proven in-reactor tests Thoria fuel has always performed comparably with UO 2, with some experiments demonstrating superior performance UNRESTRICTED / ILLIMITÉ 37

38 Conclusions Thoria-based fuels can achieve superior performance characteristics to that of UO 2 fuels, provided pellet fabrication technologies are used to achieve a high quality non-granular microstructure UNRESTRICTED / ILLIMITÉ 38

39 Critical Experiments at AECL UNRESTRICTED / ILLIMITÉ 39

40 Critical Experiments AECL has a long history of critical experiments involving thorium fuels Three sets of experiments, dating back to the 1960 s HEU/Th Pu/Th U-233/Th Performed in the ZED-2 (Zero Energy Deuterium) reactor at Chalk River Laboratories UNRESTRICTED / ILLIMITÉ 40

41 HEU/Th (1966) 98.5% ThO2, 1.5% HEU (93% U-235) 19-element bundles 7 test channels U-233/Th (1991) 98.6% ThO2, 1.4%UO2, (97.6% U-233) 36-element bundles 7 test channels Pu/Th (1986) 36-element bundles 97.8% ThO2, 2.2% PuO2 (1.8% fissile) 7 test channels UNRESTRICTED / ILLIMITÉ 41

42 Substitution Experiments Requires only 35 bundles substituted in a reference lattice compared to about 275 bundles for a critical core Can measure void-reactivity and lattice reactivity for fuel/coolant temperatures in the range 25 to 300 o C Substituted Channels 28-Element Reference Lattice UNRESTRICTED / ILLIMITÉ

43 Physics Experiments in ZED-2 Substitution Experiments determine fuel properties (buckling/reactivity) when only a limited amount of fuel (typically seven assemblies) is available Flux Maps copper foils are irradiated to measure the flux shape and derive extrapolation distances Reaction Rate (Fine Structure) - provide detailed information about neutron distributions (in space and energy) in and around a fuel channel, as well as fission-rate and conversion ratio data within the fuel. Used for qualification of reactor physics codes program ongoing UNRESTRICTED / ILLIMITÉ 43

44 Conclusions AECL has a long history of critical experiments in the ZED-2 facility These are substitution experiments, with 7 channels of the test fuel Wide variety of experiments have been performed Different lattice pitches Different coolants Heated channels, etc These experiments are currently being analysed as part of a program to qualify physics codes for design of thorium fuel cycles UNRESTRICTED / ILLIMITÉ 44

45 Homogeneous Thorium Fuel Cycles in CANDU Reactors Bronwyn Hyland Global 2009 September 10, 2009 UNRESTRICTED / ILLIMITÉ

46 Overview Motivation Calculation Fuel Design Results Low and high burnup Pu driven once-through Low and high burnup Pu driven with U-233 recycle Low and high burnup LEU driven once-through Conclusions UNRESTRICTED / ILLIMITÉ

47 UNRESTRICTED / ILLIMITÉ Thorium Fuel Configurations

48 Thorium Fuel Cycles The simplest implementation of a thorium-based fuel is in a homogeneous thorium fuel cycle. The CANDU reactor can efficiently exploit thorium in a homogeneous thorium fuel cycle (a small amount of fissile material can go a long way) The introduction of U-233 recycle can make a dramatic improvement in fissile utilization UNRESTRICTED / ILLIMITÉ

49 Calculation Lattice cell calculations performed with WIMS-AECL 6 cases studied: Fissile Driver Once- Through/Recycle Pu Once-Through 20 Burnup (MWd/kg) 45 Recycle LEU Once-Through UNRESTRICTED / ILLIMITÉ

50 Calculation Models developed to maximize the amount of energy derived from thorium Report results here on: Exit burnup Fuel temperature coefficient Maximum linear element rating Percentage of energy derived from thorium Distribution of U-233 and Pa-233 in the bundle UNRESTRICTED / ILLIMITÉ

51 Fuel Design High burnup and recycle cases the fuel was graded Reduce size, increase number of fuel pins to decrease linear element ratings Centre pin of zirconia-filled Hf Centre Inner Intermediate Outer UNRESTRICTED / ILLIMITÉ

52 Fuel Design Case Burnup Bundle average Pu wt% or LEU wt% Pu-driven, OT Bundle average U-233 wt% Low 3.5 N/A High 4.9 N/A Pu-driven, Recycle Low High LEU-driven Low 12.2 N/A High 14.2 N/A UNRESTRICTED / ILLIMITÉ

53 Results Case Pu-driven, OT Pu-driven, Recycle Burnup (MWd/kg) FTC (μk/ºc) Max. LER (kw/m) % Energy from Th LEU-driven UNRESTRICTED / ILLIMITÉ

54 U Pa-233 U Pa Th-232 U Pa-233 U Pa Th-232 Pu-Driven, U-233 Recycle MWd/kg 45 MWd/kg Burnup (MWd/kg) Burnup (MWd/kg) Inner Ring Outer Ring Intermediate Ring Total UNRESTRICTED / ILLIMITÉ

55 % of Total Fissions % of Total Fissions Pu-Driven, U-233 Recycle MWd/kg 45 MWd/kg Burnup (MWd/kg) Fissions from Pu239 and Pu241 Fissions from Th232, U233, and U Burnup (MWd/kg) UNRESTRICTED / ILLIMITÉ

56 Conclusions CANDU reactor can exploit homogeneous thorium fuel cycles Low BU Pu-driven case gives the best result for energy from thorium, more energy proportionally required from driver fuel for higher burnup For once through cases higher burnup gives higher energy from thorium Maximum energy from thorium corresponds to minimum poison in the centre pin Results in grading of fissile Constrained by LER UNRESTRICTED / ILLIMITÉ

57 CANDU Reactor Heavy Water Moderator Good neutron economy On-power Fuelling Calandria Tube Simple fuel bundle CANDU fuel channel Pressure Tube UNRESTRICTED / ILLIMITÉ

58 ZED-2 Reactor ZED-2 : Zero Energy Deuterium, successor to ZEEP Low-power (200 w), heavy-water moderated reactor Tank-type (3.36 meter diameter, 3.35 meter high) Peak flux of 1x10 9 n/cm 2 /sec Designed for CANDU reactor support First criticality in September 1960 Reactor control is via moderator level adjustment Primary research activity is support of reactor physics code development for CANDU reactors UNRESTRICTED / ILLIMITÉ 58

59 Top Shield Doors Moveable Beam Experimental Fuel Rods Heavy Water Moderator Side Shield Doors Graphite Reflector Air Duct Hoist Heavy Water Dump Tanks Cross-Section of ZED-2 Aluminum Tank (Calandria) Shielding Control Room Heavy Water Pump Dump Valves Filling Valves Drain Valves These valves control the heavy water level in the UNRESTRICTED / ILLIMITÉ 59 calandria

60 Moderator Level Control Top shields Beam Chain Fuel Rods Aluminum Calandria Gap Graphite Reflector Heavy Water Dump Valve Dump Tank (1 of 3) Shut-Off and Drain Valve Fill Pump Reactor Vessel Approximately to scale 100 cm UNRESTRICTED / ILLIMITÉ 60

61 Typical ZED-2 Fuel Channel Zircoloy-4 Sheath Fuel Pressure Tube Calandria Tube Fuel Support Plate Zr-4 Channel End Plate Zr-2 Plug (in for void, out for cooled) Zr-2.5%Nb Pressure Tube Zircoloy-2 Calandria Tube ZED-2 Calandria floor UNRESTRICTED / ILLIMITÉ 61

62 U Pa-233 U Pa Th-232 U Pa-233 U Pa Th-232 Pu-Driven Once-Through MWd/kg 45 MWd/kg Burnup (MWd/kg) Inner Ring Intermediate Ring Outer Ring Total Burnup (MWd/kg) UNRESTRICTED / ILLIMITÉ

63 Fuel Temperature Coefficient (μk/ºc) Fuel Temperature Coefficient (μk/ºc) Pu-Driven Recycle, FTC MWd/kg 45 MWd/kg Burnup (MWd/kg) -8 Burnup (MWd/kg) UNRESTRICTED / ILLIMITÉ 63

64 % of Total Fissions % of Total Fissions Pu-Driven Once-Through MWd/kg 45 MWd/kg Burnup (MWd/kg) Fissions from Pu239 and Pu241 Fissions from Th232, U233, and U Burnup (MWd/kg) UNRESTRICTED / ILLIMITÉ

65 U Pa-233 U Pa Th-232 U Pa-233 U Pa Th-232 LEU-Driven, Once-Through MWd/kg 45 MWd/kg Burnup (MWd/kg) Burnup (MWd/kg) Inner Ring Outer Ring Intermediate Ring Total UNRESTRICTED / ILLIMITÉ

66 % of Total Fissions % of Total Fissions LEU-Driven, Once-Through MWd/kg 45 MWd/kg Burnup (MWd/kg) Burnup (MWd/kg) Fissions from U235, U238, Pu239, and Pu241 Fissions from U233 and Th232 UNRESTRICTED / ILLIMITÉ

67 Linear Element Rating (W/cm) Linear Element Rating (W/cm) Pu-Driven Recycle, LER MWd/kg 45 MWd/kg Inner Intermediate Outer Inner Intermediate Outer Burnup (MWd/kg) Burnup (MWd/kg) UNRESTRICTED / ILLIMITÉ 67

68 Other Solution-Based Methods Sol-gel derived clay extrusions Solution impregnation UNRESTRICTED / ILLIMITÉ 68

69 Mechanical Mixing Wet and dry processes have been evaluated Wet processes aid in the dispersion of the different powders amongst each other but require drying of the slurry danger of residual granules in pellet structure Dry processes due to the cohesive nature of ceramic-grade powders, the degree of homogeneity achieved is related to the intensity of the process used. Dusty, but no drying stage UNRESTRICTED / ILLIMITÉ 69

70 Mechanical Mixing Methods Turbula (Dry) Attrition Mill (Wet) Vibratory Mill (Dry) Homogenizer (Wet) UNRESTRICTED / ILLIMITÉ 70

Sustainability of Nuclear Power

Sustainability of Nuclear Power Dave Torgerson Senior Technical Advisor (emeritus) AECL Carleton Sustainable Energy Research Centre Seminar Series 2011 March 28 UNRESTRICTED / ILLIMITÉ 1 The drivers for