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
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 Fundamental Advantage of Heavy Water
• Heavy water has the highest moderating ratio (s/a). –Slows down neutrons with minimal absorption.
– Better than H (in H2O), 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). – 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
PT-HWR
• Operational Technology. • Future HWR variants. • Potential for further improvements. • Use R&D to find them.
UNRESTRICTED / ILLIMITÉ 5 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 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 Thorium in CANDU / PT-HWR Evolution
PT-HWR Canadian SCWR
With U-233 With U-233 Recycle Recycle Build U-233 resource U-233 + Pu U-233 + Once-through Pu Pu/Th
Innovation Once-through Once-through LEU/Th Pu/Th
Years
AECL - OFFICIAL USE ONLY / À USAGE EXCLUSIF - EACL 8 Long-term Impact
• Decay heat in spent fuel is a main parameter in determining the capacity of a long term disposal facility
1.6 Current global cycle, LWRs + HWRs 1.4 Transition to once-through thorium in CANDU Transition to fast reactors 1.2 Transition to Th with Once-through thorium gives the
1.0 same reduction as fast reactors U-233 recycle in 0.8 CANDU Once-through thorium gives 50% 0.6 reduction over current cycle 0.4
Decay heat (GW) heat Decay 0.2 0.0 0 200 400 600 800 1000 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 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 Thorium Pellet Structure
Granular
Homogeneous
UNRESTRICTED / ILLIMITÉ 11 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 MWe prototype PT-HWR.
• Pure ThO2, (U,Th)O2 , and (Pu,Th)O2 • NRU still operational.
Irradiation # of experiments Irradiation Time Facility frame NPD 1 1976 NRX 20 1962-1992 NRU 28 1966-2005 WR1 18 1970-1980
UNRESTRICTED / ILLIMITÉ 12 NRX, NRU, WR-1, NPD
• NPD-2 • WP-1 • NRU • NRX
UNRESTRICTED / ILLIMITÉ 13 Critical Experiments
• AECL: long history of critical experiments involving thorium-based fuels. • Three sets of experiments, dating back to the 1960’s – HEU/Th (1966-1968) – 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 Alternative Fuel Bundle and Core Design Options
Heterogeneous, Homogeneous 1.4 Mixed Bundle Bundle 1.2
1 NU 0.8
0.6
0.4 CANDU
0.2 Checkerboard Annular Seed- Blanket Seed-Blanket 0 Cores Cores 1 2 3 4 5 6 7 8 9 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 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 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 0 B B 0 0 0 0 0 B B B S S S S S S B B B 0 0 0 0 0 B Checkerboard Core Designs C 0 0 0 0 B S S B S S B B S S B S S B 0 0 0 0 C C 0 0 0 0 B S S S S S S S S S S S S B 0 0 0 0 C
D 0 0 0 B S S B B S S B B S S B B S S B 0 0 0 D D 0 0 0 B S S S S S S S S S S S S S S B 0 0 0 D Fissile Utilization, Relative to Natural Uranium Uranium Natural to Relative Utilization, Fissile E 0 0 B S S B S S B B S S B B S S B S S B 0 0 E E 0 0 B S S S S S S S S S S S S S S S S B 0 0 E Annular Core Designs F 0 0 B S B B S S B B S S B B S S B B S B 0 0 F F 0 0 B S S S S S S S S S S S S S S S S B 0 0 F
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 M B S S S S S S S S S S S S S S S S S S S S B M Hafnium Tube 3.8% Pu N B S S S B B S S B B S S B B S S B B S S S B N N B S S S S S S S S S S S S S S S S S S S S B N 3 mm thick
O B S S S B B S S B B S S B B S S B B S S S B O O B S S S S S S S S S S S S S S S S S S S S B O 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 Q 0 B S S S S S S S S S S S S S S S S S S B 0 Q PT R 0 0 B S B B S S B B S S B B S S B B S B 0 0 R R 0 0 B S S S S S S S S S S S S S S S S B 0 0 R Zr Rod S 0 0 B S S B S S B B S S B B S S B S S B 0 0 S 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 0 0 0 B S S S S S S S S S S S S S S B 0 0 0 T CT T 0 0 0 B S S B B S S B B S S B B S S B 0 0 0 T U 0 0 0 0 B S S S S S S S S S S S S B 0 0 0 0 U U 0 0 0 0 B S S B S S B B S S B S S B 0 0 0 0 U V 0 0 0 0 0 B B B S S S S S S B B B 0 0 0 0 0 V V 0 0 0 0 0 B B B S S S S S S B B B 0 0 0 0 0 V W 0 0 0 0 0 0 0 0 B B B B B B 0 0 0 0 0 0 0 0 W W 0 0 0 0 0 0 0 0 B B B B B B 0 0 0 0 0 0 0 0 W 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 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 Moderator
AECL - OFFICIAL USE ONLY / À USAGE EXCLUSIF - EACL 15 Potential to Increase Utilization
• Achieve ~ 20% to 100% higher utilization of fissile fuel than PT-HWR with NU fuel in an OTT cycle.
2.2
2.0
1.8
NU -
1.6
HWR - 1.4
1.2 PT-HWR NU 35-LEU/Th - 8-Th 1.0 35-LEU/Th 35 LEU/Th - ZrO2 Rod 0.8 21-LEU/Th 0.6 21 LEU/Th - ZrO2 Rod 35-Pu/Th-8-Th 0.4 35-Pu/Th
Fissile Utilization Relative to PT to Relative Fissile Utilization 35-Pu/Th - ZrO2 Rod 0.2 21-Pu/Th 21-Pu/Th - ZrO2 Rod 0.0 0.000 0.005 0.010 0.015 0.020 0.025 0.030 0.035 0.040 0.045 0.050 0.055 0.060 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~0.9 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.
UNRESTRICTED / ILLIMITÉ – Recycle in future. 17
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 More Info
• Visit:
• http://www.aecl.ca/site3.aspx
• https://canteach.candu.org/Pages/Welcome.aspx
• http://cns-snc.ca/home
UNRESTRICTED / ILLIMITÉ 19 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 UNRESTRICTED / ILLIMITÉ 21 Fast Fission in Fertile Isotopes
• U-238, Th-232
U-238
Th-232
UNRESTRICTED / ILLIMITÉ 22 AECL Work on ADS / HFFR for U-233 Production from Th-232 • 1962-1982 - 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 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 Neutron Production in Fissile Isotopes • Variation of neutron production per neutron absorted. Isotope Thermal Spectrum** Fast Spectrum***
U-233 2.28 - 2.30 2.31 – 2.42
U-235 2.03 - 2.07 1.93 - 2.17
Pu-239 1.80 - 2.11 2.49 - 2.68
Pu-241 2.14 - 2.15 2.72
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 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 UO2 and in some cases demonstrated superior performance
UNRESTRICTED / ILLIMITÉ 26 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 Fabrication – Sol-Gel
• Microspheres from 15 to > 1000 μm
UNRESTRICTED / ILLIMITÉ 28 Experiment Summary Table
Irradiation # of Irradiation Facility experiments Time frame NPD 1 1976
NRX 20 1962-1992
NRU 28 1966-2005
WR1 18 1970-1980
Each experiment consisted of a series of irradiations Other irradiations were done and reported in the literature
UNRESTRICTED / ILLIMITÉ 29 Granular Pellet Structure
95% Dense
UNRESTRICTED / ILLIMITÉ 30 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 MWe power reactor
• Pure Thoria, Thoria-UO2 & Thoria-PuO2
UNRESTRICTED / ILLIMITÉ 31 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 Six Thoria-PuO2 Bundles (1)
• Irradiation performed in NRU • 36 element Bruce type fuel bundle 86.05
wt% Th and 1.53 wt% Pu in (Th, Pu)O2 • Objectives
–To verify the ability of (Th, Pu)O2 fuel to operate at significant power outputs to burnups of 42 MWd/kgHE –To examine the power-ramp performance of
Zircaloy clad (Th, Pu)O2 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-PuO2 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 boundary which limit fuel performance • Low gas release • Low sheath strain
• Significant % of PuO2 present as agglomerates
Outer element in Bundle ADC-1
UNRESTRICTED / ILLIMITÉ 34 Thoria Demonstration Irradiation PIE
• Higher than expected gas release due to granular structure of pellets (WR1-1007 tests)
Granules
UNRESTRICTED / ILLIMITÉ 35 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 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 UO2, with some experiments demonstrating superior performance
UNRESTRICTED / ILLIMITÉ 37 Conclusions
• Thoria-based fuels can achieve superior
performance characteristics to that of UO2 fuels, provided pellet fabrication technologies are used to achieve a high quality non-granular microstructure
UNRESTRICTED / ILLIMITÉ 38 Critical Experiments at AECL
UNRESTRICTED / ILLIMITÉ 39 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 • 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 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 300oC
Substituted Channels 28-Element Reference Lattice
UNRESTRICTED / ILLIMITÉ 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 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 Homogeneous Thorium Fuel Cycles in CANDU Reactors
Bronwyn Hyland Global 2009 September 10, 2009
UNRESTRICTED / ILLIMITÉ
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 Thorium Fuel Configurations
UNRESTRICTED / ILLIMITÉ 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 Calculation
• Lattice cell calculations performed with WIMS-AECL • 6 cases studied: Fissile Driver Once- Burnup Through/Recycle (MWd/kg) Pu Once-Through 20
45
Recycle 20
45
LEU Once-Through 20
45
UNRESTRICTED / ILLIMITÉ 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 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 Fuel Design
Case Burnup Bundle average Pu Bundle average wt% or LEU wt% U-233 wt% Pu-driven, Low 3.5 N/A OT High 4.9 N/A
Pu-driven, Low 0.8 1.4 Recycle High 2.1 1.4
LEU-driven Low 12.2 N/A
High 14.2 N/A
UNRESTRICTED / ILLIMITÉ Results
Case Burnup FTC Max. LER % Energy (MWd/kg) (μk/ºC) (kW/m) from Th Pu-driven, 19 -3.8 56 19 OT 45 -5.0 61 29
Pu-driven, 20 -7.5 49 78 Recycle 44 -7.3 59 66
LEU-driven 20 -12.7 51 25
44 -10.7 60 41
UNRESTRICTED / ILLIMITÉ Pu-Driven, U-233 Recycle
20 MWd/kg 45 MWd/kg 2.2 1.8 2.0
232 1.6
1.8 -
232
- 1.4
1.6 233 - 1.2
233 1.4 -
1.2 1.0 233 + Th + 233 1.0 - 233 + Th + 233 0.8
- 0.8 233 + Pa + 233
- 0.6
233 + Pa + 233 0.6 U - 0.4
U 0.4 233 + Pa + 233
0.2 - 0.2
233 + Pa + 233 U - 0.0 U 0.0 0 5 10 15 20 25 0 10 20 30 40 50 Burnup (MWd/kg) Burnup (MWd/kg)
Inner Ring Intermediate Ring Outer Ring Total
UNRESTRICTED / ILLIMITÉ Pu-Driven, U-233 Recycle
20 MWd/kg 45 MWd/kg
100 100 90 90 80 80 70 70 60 60 50 50 40 40 30 30 20
% of Total Fissions 20
10 Totalof Fissions% 10 0 0 0 5 10 15 20 25 0 10 20 30 40 Burnup (MWd/kg) Burnup (MWd/kg) Fissions from Pu239 and Pu241
Fissions from Th232, U233, and U238
UNRESTRICTED / ILLIMITÉ 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 CANDU Reactor
On-power Fuelling Heavy Water Moderator – Good neutron economy
Calandria Tube
Simple fuel bundle
CANDU fuel channel Pressure Tube
UNRESTRICTED / ILLIMITÉ 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 1x109 n/cm2/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 Cross-Section of ZED-2 Aluminum Tank (Calandria) Shielding Control Room Top Shield Doors Moveable Beam Experimental Fuel Rods Heavy Water Moderator Side Shield Doors
Graphite Reflector Air Duct Hoist Heavy Water Dump Tanks Dump Valves Filling Valves Drain Valves Heavy Water Pump These valves control the heavy water level in the UNRESTRICTED / ILLIMITÉ calandria 59 Moderator Level Control
Top shields
Beam Chain Fuel Rods Aluminum Calandria Gap Graphite Reflector Heavy Water
Dump Valve Dump Tank Shut-Off and Drain Valve (1 of 3) Fill Pump
Reactor Vessel Approximately to scale 100 cm
UNRESTRICTED / ILLIMITÉ 60 Typical ZED-2 Fuel Channel
Pressure Tube
Zircoloy-4 Calandria Tube Sheath Fuel Fuel Support Plate Zr-4
Channel End Plug (in for Plate Zr-2 void, out for cooled)
Zr-2.5%Nb Zircoloy-2 Pressure Tube Calandria Tube ZED-2 Calandria floor
UNRESTRICTED / ILLIMITÉ 61 Pu-Driven Once-Through
20 MWd/kg 45 MWd/kg
1.2 1.6
232 1.4
232
1.0 -
- 1.2
0.8 233
233 -
- 1.0
0.6 0.8
233 + Th + 233
233 + Th + 233 -
- 0.4 0.6
233 + Pa + 233 233 + Pa + 233 - 0.4
- 0.2 U U 0.2
0.0
233 + Pa + 233
233 + Pa + 233 -
- 0 5 10 15 20 0.0 U U Burnup (MWd/kg) 0 20 40 60 Burnup (MWd/kg) Inner Ring Intermediate Ring Outer Ring Total
UNRESTRICTED / ILLIMITÉ Pu-Driven Recycle, FTC
20 MWd/kg 45 MWd/kg 0 0 0 10 20 30 0 20 40 60 -1 -1 -2 -2 -3 -4 -3
-5
- 4
k/ºC) -6
μ -5
k/ºC)
( μ
-7 ( -6 -8
-9 -7 Fuel Temperature Coefficient Coefficient Temperature Fuel
-10 Coefficient Temperature Fuel -8 Burnup (MWd/kg) Burnup (MWd/kg)
UNRESTRICTED / ILLIMITÉ 63 Pu-Driven Once-Through
20 MWd/kg 45 MWd/kg 100 100 90 90 80 80 70 70 60 60 50 50 40 40
30 30 % of Total Fissions Total of % % of Total Fissions 20 20 10 10 0 0 0 5 10 15 20 0 10 20 30 40 50 Burnup (MWd/kg) Burnup (MWd/kg)
Fissions from Pu239 and Pu241 Fissions from Th232, U233, and U238
UNRESTRICTED / ILLIMITÉ LEU-Driven, Once-Through
20 MWd/kg 45 MWd/kg
1.4 1.6
1.2 1.4
232 232
- - 1.2
1.0 233
233 1.0
- - 0.8
0.8
233 + Th + 233 233 + Th + 233 -
0.6 - 0.6 233 + Pa + 233
233 + Pa + 233 0.4 0.4
- -
U U 0.2
0.2
233 + Pa + 233 233 + Pa + 233 -
- 0.0
U U 0.0 0 10 20 30 40 50 0 10 20 30 Burnup (MWd/kg) Burnup (MWd/kg)
Inner Ring Intermediate Ring Outer Ring Total
UNRESTRICTED / ILLIMITÉ LEU-Driven, Once-Through
20 MWd/kg 45 MWd/kg 100 100 90 90 80 80 70 70 60 60 50 50 40 40 30 30
20 Fissions Total of % % of Total Fissions Total of % 20 10 10 0 0 0 5 10 15 20 25 0 10 20 30 40 50 Burnup (MWd/kg) Burnup (MWd/kg) Fissions from U235, U238, Pu239, and Pu241 Fissions from U233 and Th232
UNRESTRICTED / ILLIMITÉ Pu-Driven Recycle, LER
20 MWd/kg 45 MWd/kg
60 70
50 60
50 40 Inner 40 30 Intermediate 30 Outer 20 Inner 20 10 Intermediate 10
Linear Element(W/cm) Rating Linear Outer 0 0
0 5 10 15 20 25 (W/cm) Rating Element Linear 0 10 20 30 40 50 Burnup (MWd/kg) Burnup (MWd/kg)
UNRESTRICTED / ILLIMITÉ 67 Other Solution-Based Methods
Sol-gel derived clay extrusions
Solution impregnation
UNRESTRICTED / ILLIMITÉ 68 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 Mechanical Mixing Methods
Attrition Mill (Wet) Turbula (Dry)
Vibratory Mill (Dry)
Homogenizer (Wet)
UNRESTRICTED / ILLIMITÉ 70