© SCKCEN, 2018
Spent fuel management in Belgium current status, prospects and P&T impact studies
Eef Weetjens [email protected] and Geosystems Analyses Unit
Waste and Disposal Expert Group
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2001 Catholic University of Leuven: Bio-Engineer Environmental Technology
2001- … Performance and Safety Assessments geological disposal
2016 - … (currently part-time) Support Safety Report surface disposal
Bio Eef Weetjens
Relevant activities: impact of P&T on geological disposal in the framework of
EC project RED-IMPACT
NEA P&T expert group
Development of MYRRHA (an ADS pilot facility) at SCK•CEN
National project ASOF
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Spent fuel in Belgium
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Nuclear energy in Belgium
Belgium has NPP’s at Doel (4 PWR units) and Tihange (3 PWR units)
Together they provide about half of the domestic energy production
© SCKCEN, 2018
Spent fuel origin/quantities
26/06/2019
Reactor Net capacity Start of
operation
Foreseen
lifetime
Fuel
type
Assembly
length
Burn-up
(GWd/tHM)
%U5 #
Doel 1 433 MWe 15/02/1975 50 y
UOX 8 ft 36-47-55 4.3 2359Doel 2 433 MWe 01/12/1975 50 y
Tihange 1 962 MWe 01/10/1975 50 yUOX
12 ft36-45-62
45-50-50
4.2
7.7
5109
144Doel 3 1006 MWe 01/10/1982 40 y
UOX
MOXTihange 2 1008 MWe 01/02/1983 40 y
Doel 4 1033 MWe 01/07/1985 40 y
UOX 14 ft 36-44-62 4.1 3426Tihange 3 1046 MWe 01/09/1985 40 y
• Values for MOX in red
• Reference values in green44.88 4.2
© SCKCEN, 2018
Spent fuel (SF) management in the past
<1993: Reprocessing & reuse of Pu (and REPU) as MOX fuel
>1993: no more reprocessing
SF & waste projections at reactor end of life
630 tHM irradiated fuel reprocessed
66 tHM irradiated MOX fuel
390 canisters (150l) vitrified high-level waste HLW
432 canisters (150l) with compacted hulls and endpieces
4643 tHM irradiated UOX fuel
SF management options
Direct disposal
Reprocessing (partial/full)
Reprocessing and advanced partitioning for conditioning (P&C) and/or transmutation (P&T)
Former reprocessing activities and projections of nuclear waste
Ultimate waste
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Belgian disposal concepts
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Categorisation of nuclear waste in Belgium
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ONDRAF/NIRAS planned (2021) surface disposal facility for category A waste
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!! Radiological limits, Cs and Sr waste
• activity concentration at different scales (waste drum, monolith, module, groups of modules and tumulus)
• total activity limitations for the disposal facility
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ONDRAF/NIRAS proposed geological repository for disposal of B & C waste
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Shielded waste packages: supercontainers and monoliths
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Supercontainer
Monolith-B
MOX SF UOX SF HLW
Bituminised
waste
Compacted
waste
© SCKCEN, 2018
Recent changes in repository lay-out and footprint
Working hypothesis: partial reprocessing
Reprocessing all (144) MOX assemblies, and
Reprocessing roughly about 1000 assemblies of each of the UOX (8ft, 12ft and 14ft) type fuel assemblies
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© SCKCEN, 2018
1. Partitioning of (activation- and) fission products for transmutation (P&T)
or (interim) storage and conditioning in dedicated matrices (P&C)
Long-lived: 99Tc (214ky), 126Sn (230ky), 79Se (356ky), 93Zr (1.53My), 135Cs (2.3My), 107Pd (6.5My) and 129I (16.1My)
only 99Tc en 129I are theoretically fit for transmutation, but efficient transmutation is hard to achieve
Heat producing: 137Cs (30y) and 90Sr (29y)
Removal of Mo and noble metals higher glass loading
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Emphasis on optimisation of
repository footprint
P&T as a means to reduce the waste burden: FP
Emphasis on reduction of RN lifetime
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2. Separation of actinides to recycle/transmute
still high amount of fissile/fertile materials after SF irradiation
93,6% U
1,0% Pu
0,08% Np
0,18% Am
0,002% Cm
Rest: ~5% fission and activation products
At industrial level: in advanced reactor types with fast neutrons
generation IV reactors: critical reactors
ADS (Accelerator Driven Systems): sub-critical reactors
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To make optimal use of fissile resources
To reduce radiotoxicity
P&T as a means to reduce the waste burden: actinides
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Impact of P&T on waste disposal
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EC RED-IMPACT project (2004-2007)/(NEA EG P&T (2006))
A1: reference “once through” cycle in PWRsUOX spent fuel
A2: mono-recycling of Pu as MOX in PWRsV-HLW, ILW, MOX spent fuel
A3: multi-recycling of Pu in (Na-cooled) FRsV-HLW, ILW
B1: multi-recycling of Pu and MA in (Na-cooled) FRsV-HLW, ILW
B2: double strata cycle of PWR’s en ADSV-HLW (UOX, MOX, ADS), ILW (MOX, ADS-pyro, ADS-oper)
open
cycle
partially
closed
cycle
closed
cycle
All of these are ‘equilibrium’ scenarios
Industrial fuel cycles
Innovative fuel cycles
© SCKCEN, 2018
Impact on natural U use
Natural U use for production of 1 TWh(e):
B2: 25% more efficient w.r.t. U consumption
Fast reactors (A3/B1): efficiency x 100 and more through use of natural or depleted U in
MOX instead of enriched U
Fuel cycle A1 A2 A3 B1 B2
Nat. U consumption kg/TWh(e) 20723 18448 986 106 15766
normalised 1 0.89 0.048 0.0051 0.76
A1: open cycle PWR with UOX fuelA2: mono-recycling of Pu as MOX in PWRsA3: multi-recycling of Pu in Na-cooled FRsB1: multi-recycling of Pu and MA in Na-cooled FRsB2: double strata cycle of PWR’s and ADS
© SCKCEN, 2018
Impact on radiotoxicity to be disposed
Radiotoxicity /10
if Pu is recycled
(multi-recycling)
Radiotoxicity /100
if Pu and MA are
recycled
Radiotoxicity = activity × dosefactor ingestion
© SCKCEN, 2018
Impact on long-term dose for repository in clay
Typical bimodal shape:
actinides are very well sorbed
in clay host rocks
Differences mainly due to
fate of I-129, and amounts of
ILW produced
© SCKCEN, 2018
Impact on waste volumes
Dimensions of waste packages in Red-Impact (not much more than an overpack)
Dimensions of ONDRAF/NIRAS supercontainers/monoliths
Fuel cycle A1 A2 A3 B1 B2
TOTAL HLW (m3/TWhe) 3.86 2.13 1.27 1.21 1.41
relative TOTAL HLW (-) 1.00 0.55 0.33 0.31 0.37
TOTAL HLW + ILW (m3/TWhe) 3.86 4.62 6.57 6.50 4.75
relative HLW +ILW (-) 1.00 1.20 1.70 1.68 1.23
Fuel cycle A1 A2 A3 B1 B2
TOTAL HLW (m3/TWhe) 27.01 22.85 15.82 14.96 17.53
relative TOTAL HLW (-) 1.00 0.85 0.59 0.55 0.65
TOTAL HLW + ILW (m3/TWhe) 27.01 27.25 25.19 24.33 23.44
relative HLW +ILW (-) 1.00 1.01 0.93 0.90 0.87
© SCKCEN, 2018
Impact on repository footprint
Theoretical maximum disposal density: decay heat calculations versus near field temperature criterion <100°C
Variants of B1: Impact of separation of 137Cs and 90Sr:
Cs and Sr streams are individually vitrified (waste loading 60%)
100 years decay storage
Fuel cycle A1 A2 A3 B1 B2
TOTAL HLW (m2/TWhe) 711 464 174 94 145
relative (-) 1.00 0.65 0.24 0.13 0.20
! No ILW considered
Fuel cycle B1.1 (40FP-60Cs-60Sr) B1.4 (60FP-60Cs-60Sr)
TOTAL HLW (m2/TWhe) 21.86 21.95
relative (-) 0.031 0.031
Factor ~10
Factor ~30
© SCKCEN, 2018
Application to Belgian geological repository: impact on gallery length (km)
MA+FP P&T case based on extrapolations from Oigawa et al. 2006
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Application to Belgian geological repository: impact on footprint (km2)
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Conclusions from impact studies so far
Geological disposal is needed in every scenario considered.
The time needed for isolation & confinement in geological disposal is equal in all
scenarios, because it depends on impact of mobile fission and activation
products, which are not targeted in any P&T scenario
Partitioning helps to reduce repository size
Full reprocessing: ↆ needed gallery length with factor 2
FP Partitioning (Cs/Sr decay): ↆ needed gallery length with factor 5
Transmutation helps to reduce the waste’s radiotoxicity
Pu multi-recycling: ↆ radiotoxicity with factor 10
Pu multi-recycling + MA transmutation: ↆ radiotoxicity with factor 100
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Ongoing studies and projects
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SCK•CEN’s MYRRHA Project: an Accelerator Driven System
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REACTOR
BUILDINGLINAC
FRONT-
END
~ 500 m
~ 200 m
Masterplan full Myrrha
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Key technical objective of the MYRRHA project: an Accelerator Driven System
MYRRHA – An Accelerator Driven System
Demonstrate the ADS concept at pre-industrial scale
Fast neutron source multipurpose and flexible
irradiation facility
Accelerator
particles protons
beam energy 600 MeV
beam current 2.4 to 4 mA
Reactor
power 65 to 100 MWth
keff 0,95
spectrum fast
coolant LBE
Target
main reaction spallation
output 2·1017 n/s
material LBE (coolant)
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MYRRHA application portfolio
Radio-isotopes
Fundamental
research
Multipurpose
hYbrid
Research
Reactor for
High-tech
Applications
Fission GEN IV Fusion
Source: SCK•CEN MYRRHA Project Team, MYRRHA Business Plan
SMR LFR
SF/ Waste
© SCKCEN, 2018
Governmental support for MYRRHA / phased approach
Belgium allocated 558 MEUR for 2019 – 2038:
Phase 1
287 MEUR investment (CapEx) for building MINERVA (Accelerator up 100 MeV + PTF) for
2019 – 2026
156 MEUR for OpEx of MINERVA for the periode 2027-2038
Phases 2-3
115 MEUR for further design, R&D and Licensing for 2019-2026
A stage-gate decision will be taken in 2026 whether to proceed with phases 2
and 3, either sequentially, or in parallel
Belgium requests to establish an International non-profit organization
(AISBL/IVZW) in charge of the MYRRHA facility for welcoming the international
partners
© SCKCEN, 2018
Governmental support for energy transition: the ASOF project
Advanced Separation for Optimal management of spent Fuel (2018-2022)
targets the development of new, innovative processes for the separation (WP1), conversion
(WP2) and conditioning (WP3) of spent fuel
WP1: MA separation (Am) and separation of short-living FP (Cs, Sr)
WP2: conversion of Am concentrate to oxide for production of Am transmutation targets
T2.1 study of production stability of the sol-gel and downblending processes through
simulant materials (see next slides and sciencedirect)
© SCKCEN, 2018
Manual microsphere fabrication
33
UO2(NO3)2 solution
1.) sol preparation
Ln(NO3)3 solution
Ln = Nd(III), Ce(III)
HMTA/urea solution
Metal blend
Metal blend
Sol silicone oil
ϑ = 90 °C
2.) gelation
1. petrolium ether (2x)
2. ammonia, w(NH3) = 12.5 %, (3x)
3.) washing and aging
4.) drying at air
5.) thermal treatment
© SCKCEN, 2018
Internal gelation: Gelification and post gelation treatment
34
washing
aging
before drying
after dryingdroplets during gelation
© SCKCEN, 2018
Pure U and U/Nd particle preparation
35
after gelation washing aging before drying after drying
U particles
χ(Nd) = 20 %
R(HMTA) = R(urea) = 1.2 𝑅 𝑥 =𝑛(𝑥)
𝑛(𝑀𝑛+)
U/Nd particles
© SCKCEN, 2018
Pure U vs. U/Ce particle preparation
36
after gelation washing aging before drying after drying
U particles
χ(Ce) = 5 %
precursor:
Ce(NO3)3∙6H2O)
R(HMTA) = R(urea) = 1.2 𝑅 𝑥 =𝑛(𝑥)
𝑛(𝑀𝑛+)
U/Ce particles
© SCKCEN, 2018
Preparation of U/Ce particles using Ce4+ as precursor
37
5 % Ce: succeeded
10 % Ce: agglomeration in
gelation column
© SCKCEN, 2018
Governmental support for energy transition: the ASOF project
Advanced Separation for Optimal management of spent Fuel (2018-2022)
targets the development of new, innovative processes for the separation (WP1), conversion
(WP2) and conditioning (WP3) of spent fuel
WP1: MA separation (Am) and separation of short-living FP (Cs, Sr)
WP2: conversion of Am concentrate to oxide for production of Am transmutation targets
T2.1 study of production stability of the sol-gel and downblending processes through
simulant materials (see next slides and sciencedirect)
WP3: conditioning of waste streams from partitioning scenarios
WP4: Impact of advanced separation on a geological repository
© SCKCEN, 2018
Cs and Sr waste: category A waste?
Decay storage
Cooling period: 100 years is generally considered in the literature
To be checked if that would be enough to ‘declassify’ waste from C to B
Radiotoxicity: comparison with
disposal (capacity) limits of cat. A facility
Cs-137
– OLI (eastern tumulus): 5.57E+13 Bq
– BLI: 2.42E+14 Bq
Sr-90
– OLI (eastern tumulus): 2.20E+12 Bq
– BLI: 9.57E+12 Bq
Cs waste will also contain long-living
isotope Cs-135 (t1/2: 2.3E+6 a)
26/06/2019
time (y) Cs-137 Sr-90
0 1.34E+19 9.01E+18
30 6.70E+18 4.38E+18
60 3.35E+18 2.13E+18
90 1.68E+18 1.03E+18
120 8.38E+17 5.02E+17
150 4.19E+17 2.44E+17
180 2.09E+17 1.18E+17
210 1.05E+17 5.75E+16
240 5.24E+16 2.79E+16
270 2.62E+16 1.36E+16
300 1.31E+16 6.59E+15
330 6.54E+15 3.20E+15
360 3.27E+15 1.56E+15
390 1.64E+15 7.56E+14
420 8.18E+14 3.67E+14
450 4.09E+14 1.78E+14
480 2.05E+14 8.66E+13
510 1.02E+14 4.21E+13
540 5.11E+13 2.04E+13
570 2.56E+13 9.93E+12
600 1.28E+13 4.82E+12
630 6.39E+12 2.34E+12
660 3.20E+12 1.14E+12
© SCKCEN, 2018
General conclusions
© SCKCEN, 2018
Conclusions (1/2)
Present Belgian context:
Implementation of the law on nuclear phase-out
Suspension of reprocessing of commercial spent fuel
Belgium considers a broad spectrum of possible SF management scenarios
Direct disposal
Full (TOPMOX) or partial reprocessing
footprint /5
P&C (advanced separation to optimize waste management, with focus on
Cs/Sr separation)
P&T (decrease of radiotoxicity with focus on transmutation of Am in ADS)
footprint /10?
© SCKCEN, 2018
Conclusions (2/2)
Growing Consensus on impact of P&T:
Thermal output can be significantly reduced by partitioning of Cs/Sr (P&C) and Am (P&T)
Radiotoxicity can be significantly decreased in case of Pu (FR) and Am (ADS) recycling
No impact on long-term dose from a geological repository (determined by fission products)
No impact on already produced (conditioned) category C and B waste
New (cat. B) waste forms will be produced in P&C/ P&T scenarios
Most steps in advanced fuel cycles are in R&D phase
difficult to assess the actual benefits for GDF and for the fuel cycle as a whole
Realistic full life cycle analysis is necessary
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Thank you
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Reserve slides
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Decay heat for a reference UOX assembly
26/06/2019
© SCKCEN, 2018
Radiotoxicity (actinides) of a reference spent fuel assembly
26/06/2019
© SCKCEN, 2018
Radiotoxicity (FAPs) of a reference spent fuel assembly
26/06/2019
