SLIDE 1 Transmutation of Actinides in CANDU Reactors
- B. Hyland
- G. Dyck
- A. Morreale
- R. Dworschak
SLIDE 2 Outline
- Introduction to CANDU reactors
- Motivation for transmutation of actinides
- Transmutation of actinides in CANDU
– Group-extracted TRU in MOX – Separated Am/Cm in targets
SLIDE 3 On-power Fuelling Heavy Water Moderator – Good neutron economy CANDU fuel channel Simple fuel bundle
The CANDU Reactor
SLIDE 4
37-element bundle
SLIDE 5 Motivation
- Increase capacity
- f long-term
geological disposal
$96 billion
limited by decay heat load
SLIDE 6 What’s Contributing to the Heat Load?
FP’s at short times Actinides at long times
*Data for Russian VVER
B.R. Bergelson, A.S. Gerasimov, and G.V. Tikhomirov
SLIDE 7 Transmutation Scenarios
- Two transmutation scenarios were
examined
- Group-extracted TRU in MOX
- Separated Am/Cm in targets
SLIDE 8 TRU MOX Scenario
Cool 30 years Group extraction MOX LWR, 45 MWd/kg
calculations
calculations
for MOX
SLIDE 9 30 year cooled SNF
MOX
Initial TRU Content, g/bundle 653 Initial TRU Content, % by volume 3.3%
Initial TRU Compostion, g/kg initial TRU
Pu-238 +163 13 Pu-239
563 Pu-240
201 Pu-241 +65 30 Pu-242 +176 38 Pu Total
845 Np Total
47 Am-241
100 Am total
108 Cm total +3700 0.6 Total MA
155 Total TRU
1000 Change in actinide composition (%) at discharge burnup
SLIDE 10 MOX
0.0 20.0 40.0 60.0 80.0 100.0 120.0 140.0 160.0 180.0 200.0 10 100 1000 10000 100000
Time (years) % decay heat
Decay Heat from Actinides
Once through PWR without CANDU
SLIDE 11 Nuclide Contribution to Heat Load
Nuclide
Time Frame of Main Contribution to Heat Load % Difference
Pu-238 Less than 100 years +163 Cm-244 Less than 100 years +2641 Am-241 Less than 1000 years
Pu-239 1000-100,000 years
Pu-240 1000-100,000 years
SLIDE 12
Full-Core Calculation
Input fuel composition Lattice cell calculation: WIMS Cross-sections, Depletion Full-Core Calculation: RFSP Full-core Parameters, Dwell Time, Burnup
SLIDE 13 Am and Cm Target Channels
channels
Am and Cm in IMF 0.9% Fissile RU
SLIDE 14 Important Criteria
- Support ratio
- % transmutation
- Residence time
:
SLIDE 15
Fuel Bundle Designs
CANFLEX 43 elements 21 elements 24 elements 30 elements
SLIDE 16 2 4 6 8 10 12 14 16 18 20 22 10 20 30 40 50 60 70 80 90 100 Destruction of Americium (%) Support Ratio, GWe LWR: GWe CANDU
SLIDE 17 10 20 30 40 50 60 70 80 90 100 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Residence Time (years) % per Initial Amount of AmCm
Total Am + Cm + Pu Am + Cm All Am All Cm All Pu Am-241 Am-243 Cm-242 Cm-244 Pu-241 Pu-239 Pu-240 Pu-242
Total Am + Cm + Pu Total Am + Cm Total Am Am-241 Total Pu Total Cm
SLIDE 18 Results
- 21-element bundle
- 26% initial concentration
- Support ratio 2.5 GWe LWR : 1 GWe CANDU
- Residence time for AmCm = 5.7 years
Input kg/CANDU Exit kg/CANDU % Change Am 373 112
Cm 9 68 +700 Total Am + Cm 382 180
SLIDE 19 Full-Core Calculation
Input fuel composition Lattice cell calculation: WIMS* Cross-sections, Depletion Full-Core Calculation: RFSP Full-core Parameters, Dwell Time, Burnup
* Calculations done with a developmental version of WIMS-AECL
SLIDE 20 Summary
- CANDU reactors have unique features which
allow them to effectively transmute transuranics
– We can burn 40% of TRU – Reduce heat load by 40% at 1000 y
– We burn 70% of Am (53% or Am+Cm) – Reduce heat load by 70% at 1000 y
- Provide a significant increase in geological
repository capacity.
- Full-core calculations indicate that both fuel
cycles are feasible
SLIDE 21
SLIDE 22 30 year cooled—No Burn
1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05
Time (years) Thermal Power (W)/kg ITRU
237Np 239Np TotalNp 238Pu 239Pu 240Pu 241Pu 242Pu TotalPu 241Am 243Am TotalAm 242Cm 244Cm 245Cm TotalCm TotalTRU
SLIDE 23 Decay Heat from Actinides, MOX
1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 Time (years) Thermal Power (W)/kg ITRU 237Np 239Np TotalNp 238Pu 239Pu 240Pu 241Pu 242Pu TotalPu 241Am 243Am TotalAm 242Cm 244Cm 245Cm TotalCm TotalTRU No Burn
Once Through LWR Total TRU
SLIDE 24 The Calculation
- WIMS-AECL used to calculate neutron
fluxes
- ORIGEN-S used for the depletion
calculation
- MOX: burned to 45 GWd/t
- Assumed 3% neutron leakage
SLIDE 25
MOX, Pu Isotopes
20 40 60 80 100 120 10 20 30 40 50
Burnup (MWd/kg) % per initial TRU
238Pu 239Pu 240Pu 241Pu 242Pu TotalPu TotalTRU
SLIDE 26
MOX, Minor Actinides
2 4 6 8 10 12 14 16 10 20 30 40 50
Burnup (MWd/kg) % per intial TRU 237Np 241Am 243Am TotalAm 242Cm 244Cm TotalCm Total MA
SLIDE 27 Group Extracted TRU MOX, Cm
0.5 1 1.5 2 2.5 3 10 20 30 40 50
Burnup (MWd/kg) % per initial TRU
Total Cm 242Cm 243Cm 244Cm 245Cm
SLIDE 28
21 kg/year 90 kg/year 27 kg/year Decay for 30 years Separate Am, Cm
Am Mass Flow
63 kg/year destroyed 2.5 GWe LWR
SLIDE 29 Full-Core Results
- Avg. burnup for RU is 12.2 MWd/kg
- 3.7 channels/day, 11 bundles/day
Time- Average Refueling Ripple NU Fuel Max Channel Power (kW) 6600 7100 7300
Power (kW) 790 845 935
SLIDE 30
