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Howell [question, comment]s: Blair's transmutation of fission wastes, 02Jan2024

Table of Contents


Acronyms

ADS sub-critical fast reactors driven by a proton accelerator
ALMR fast-spectrum advanced liquid metal-cooled reactors
Am americium
BU Lattice burnup
Cm curium
DGR Deep Geological Repository
ELESTRES updated analysis codes [CBM2021 55, 56]
ENDF/B-VII.089-group nuclear data library [CBM2021 48]
FTC fuel temperature coefficient
HFR High Flux Reactor
HLW high level radioactive waste
k-infinity infinite neutron multiplication factor
LER maximum linear element rating
LLFPs long-lived fission products
LOCA transient LOCA event
LWRS conventional thermal-spectrum reactors (incl PWRs and BWRs)
PT-HWR pressure tube heavy water reactor
RFSP 3.5.1 Time-average equilibrium core physics & refuel two-group, 3D (x,y,z) diffusion code [CBM2021 51]
SERPENT alternative lattice physics code and associated nuclear data set [CBM2021 46]
TRU transuranic elements = Np + Pu + Am + Cm, Am + Cm, or just Am
WIMS-AECL Version 3.1 Lattice physics modeling [CBM2021 47]
WIMS-AECL WIMS Utilities Version 2.0 Lattice physics data [CBM2021 50]

CBM = Colton, Bromley Mar2021


Colton, Bromley Jul20

Acronyms

ADS sub-critical fast reactors driven by a proton accelerator
ALMR fast-spectrum advanced liquid metal-cooled reactors
Am americium
BU Lattice burnup
Cm curium
DGR Deep Geological Repository
ELESTRES updated analysis codes [CBM2021 55, 56]
ENDF/B-VII.089-group nuclear data library [CBM2021 48]
FTC fuel temperature coefficient
HFR High Flux Reactor
HLW high level radioactive waste
k-infinity infinite neutron multiplication factor
LER maximum linear element rating
LLFPs long-lived fission products
LOCA transient LOCA event
LWRS conventional thermal-spectrum reactors (incl PWRs and BWRs)
PT-HWR pressure tube heavy water reactor
RFSP 3.5.1 Time-average equilibrium core physics & refuel two-group, 3D (x,y,z) diffusion code [CBM2021 51]
SERPENT alternative lattice physics code and associated nuclear data set [CBM2021 46]
TRU transuranic elements = Np + Pu + Am + Cm, Am + Cm, or just Am
WIMS-AECL Version 3.1 Lattice physics modeling [CBM2021 47]
WIMS-AECL WIMS Utilities Version 2.0 Lattice physics data [CBM2021 50]

CBM = Colton, Bromley Mar2021


Colton, Bromley Jul2018 mixed oxide thorium based fuels

Ashlea Colton, Blair Bromley Jul2018 "Lattice physics evaluation of mixed oxide thorium based fuels for use in pressure tube heavy water reactors" Annals of Nuclear Energy 117(2):259-276, https://www.researchgate.net/publication/326100919_Lattice_physics_evaluation_of_35-element_mixed_oxide_thorium-based_fuels_for_use_in_pressure_tube_heavy_water_reactors


ABSTRACT
A series of 2-D lattice physics depletion studies were carried out as part of conceptual scoping studies to evaluate thorium-based fuels in Heavy Water Reactors. Fuel bundles used for the study consisted of 35 fuel elements arranged in two outer rings comprised of thorium mixed with a fissile driver of either reactor grade plutonium (3.5-4.5 wt% PuO 2 ), low enriched uranium (5 wt% 235 U/U, 40-50 wt% LEUO 2 ) or uranium-233 (1.8 wt% 233 UO 2 ). Lattice physics- based estimated for fuel temperature coefficients and coolant void reactivity values were found to be lower than those for conventional natural uranium HWR fuel. The low burnup LEU/thorium fuel bundle shows the most promise as a feasible means of generating U-233 in a PT-HWR from a safety and fuel performance perspective.

Howell [question, comment]s :
- none for the moment. I am unfamiliar with many of the key [measurements, criteria] and they normally would be, so I would have to put in much more time to understand the significance of results. Sounds like one can operate current equipment with Thorium safely, and reasonable output?


Bromley, Alexander Dec2018 thorium and depleted uranium in sub-critical GC-PTR

Blair Bromley, Jude Alexander Dec2018 "Assessment of fast-spectrum blanket lattices for breeding fissile fuel from thorium and depleted uranium in an externally driven sub-critical gas-cooled pressure tube reactor" CNL Nuclear Review 7(2):201-220, DOI:10.12943/CNR.2018.00010 https://www.researchgate.net/publication/329346351_ASSESSMENT_OF_FAST-SPECTRUM_BLANKET_LATTICES_FOR_BREEDING_FISSILE_FUEL_FROM_THORIUM_AND_DEPLETED_URANIUM_IN_AN_EXTERNALLY_DRIVEN_SUB-CRITICAL_GAS-COOLED_PRESSURE_TUBE_REACTOR

ABSTRACT
To ensure long-term nuclear energy security, it is advantageous to consider the use of externally driven sub- critical systems for producing fissile fuel to supply fleets of thermal-spectrum reactors as an alternative to using fast- spectrum or thermal-spectrum breeder reactors.

Computational/analytical neutronics and heat transfer studies have been carried out for gas-cooled fuel bundle lattices with mixtures of fertile thorium and depleted uranium (DU) that could be used in the blanket region of a sub-critical fast reactor driven either by a fusion reactor in a hybrid fusion-fission reactor (HFFR) system, or an accelerator-based spallation neutron source in an accelerator driven system (ADS). The HFFR or ADS concept envisioned is one with a simple cylindrical geometry. The annular-cylindrical blanket is approximately 10 m long, can be made 2–5 m thick (1.0 m ≤ R blanket ≤ 3.0 m to 6.0 m), and is filled with a repeating square lattice of pressure tubes filled with 0.5 m long fuel bundles that are made of (DU,Th)O 2 , with various mixtures of Th and DU, and refuelled periodically online. Although using blankets made of pure DUO 2 or ThO 2 are viable options to analyze, mixing DUO 2 with ThO 2 can help alleviate any potential proliferation concerns, since any 233 U produced from breeding will be denatured by the presence of 238 U in (DU, Th)O 2 .

Lattice calculations demonstrate that the total fissile content in the fuel after an extended period of burnup (50 MWd/kg) will be approximately the same, regardless of the mixture of DU and thorium used, and that the content of americium and 232 U in the irradiated fuel will be <0.01 wt%/initial heavy metal.


Howell - same abstract as for Colton, Bromley Jul2018???


Colton, Bromley Mar2021 PT Heavy Water Reactor to Destroy Americium and Curium

Ashlea Colton, Blair Bromley Mar2021 "Reactor Physics Assesalternative lattice physics code and associated nuclear data set, such as SERPENT [61 ]sment of Potential Feasibility of Using Advanced, Nonconventional Fuels in a Pressure Tube Heavy Water Reactor to Destroy Americium and Curium" March 2021 Nuclear Technology 207(1):1-23, DOI:10.1080/00295450.2020.1853466, https://www.researchgate.net/publication/350481335_Reactor_Physics_Assessment_of_Potential_Feasibility_of_Using_Advanced_Nonconventional_Fuels_in_a_Pressure_Tube_Heavy_Water_Reactor_to_Destroy_Americium_and_Curium

Table I. Expected Time Scales and Thermal Neutron Fluxes Required to Destroy MAs and non-Fissile Isotopes of Pu
Actinide Half Life (years) Dominant Decay Mode Years to Decay naturally
to 1.0×10^-4 n/cm^2/s
*_ 		Thermal σ_absorb (barns) Thermal σ_fission (barns) Fission Spectrum σ_fission (barns) Years to Reduce to
1.0×10^-4 n/cm^2/s
with
1.0×10^+14 n/cm^2/s
Neutron Flux
**		 Flux to Reduce to
1.0×10^-4 n/cm^2/s
within 100 years
**
Np-236 154,000 EC to U-236*** 2,046,742 3073.9 2,453 1.92 1.0 9.50E+11
Np-237 2,144,000 Alpha to Pa-233 28,494,906 144 0.019 1.34 20.3 2.03E+13
Pu-238 87.7 Alpha to U-234 1,166 473.2 15.18 1.99 6.1 5.64e+12
Pu-240 6564 Alpha to U-236 87,239 263.7 0.053 1.36 11.1 1.11e+13
Pu-242 373,300 Alpha to U-238 4,961,357 16.8 0.0023 1.13 173.6 1.74e+14
Am-241 432 Alpha to Np-237 5,744 534.7 2.711 1.38 5.5 5.37e+12
Am-242m 141 IT to Am-242*** 1,874 7460. 6235 1.84 0.4 3.71e+11
Am-243 7,370 Alpha to Np-239 97,951 70.6 0.103 1.08 41.4 4.13e+13
Cm-243 29 Alpha to Pu-239 387 662.8 549.6 1.94 4.4 3.27e+12
Cm-244 18 Alpha to Pu-240 241 14.3 0.906 1.57 110.6 1.2e+14
Cm-245 8,500 Alpha to Pu-241 112,970 1961.4 1674 1.74 1.5 1.49e+12
Cm-246 4,760 Alpha to Pu-242 63,263 1.3 0.124 1.23 2212 2.29e+15
Cm-247 15,600,00 00 Alpha to Pu-243 207,332,337 121. 70.8 1.91 24.1 2.41e+13
Cm-248 348,000 Alpha to Pu-244 4,625,106 2.6 0.328 1.25 1,118 1.12E+15

*_ Calculated using equation N(t)/N(0) = exp(-1.0*0.693/(T_half*t))
** Calculated using equation N(t)/N(0) = exp(-1.0*0.693/(T_half*t) + σ*φ*t)).
*** Decay mode notes:
Fig 5 Net Change in Mass of U, Pu, and MAs in Seed and Blanket Bundle Concepts at Exit Burnup

Howell [question, comment]s :
Are [Am, Cm] already present in the feed fuel at very low levels, or are they produced during PT-HWR operation, either from the fuel or reactor materials of construction? The paper addresses the following issues after the introductory part :
What are the comparative costs associated with :
  • [extraction, concentration] of [Am, Cm] from spent fuel
  • running the transmutation of the [Am, Cm] target bundles in a PT-HWR reactor (perhaps for years to get acceptable levels)?
Does this require a dedicated PT-HWR, or can the [Am, Cm] bundles be placed with normal fuel in a normal CANDU over decades until threshold levels are obtained?
p02h0.45 Abstract "The potential to achieve net zero production of Am and Cm in a single thermal-spectrum reactor, such as a PT-HWR, could help eliminate the need to build and qualify a Deep Geological Repository (D
p03h0.40 Introduction "Long-term isolation and storage of MAs in a Deep Geological Repository (DGR) to ensure the protection of the environment for millions of years is considered scientifically and technologically feasible [1], [2]. However, it may be preferable to separate and destroy Am and Cm through direct fission, or transmutation into fissile isotopes by neutron capture, followed by fission."
Howell: Bernie Gorsky of NRCan/CANMET Mining in Ottawa (the divisions and branches have changed names a lot) did testing of core samples for DGRs. I was ignorant of the problems, but he took time to get me to realise that every time you create an openeing in rock formations, there is wall damage. Natural processes can also do that. As I understood it >10-20 years ago, the thinking was to store material in sturdy containers in DGRs, so that it could be removed perhaps decades later for reprocessing, or moving to a better site. Getting rid of it is nice if affordable - my occasional excursions into themes of history suggest that [political systems, societies] don't last forever either (normally, perhaps more like decades to a couple of centuries?).
p04h0.00 Table I Expected Time Scales and Thermal Neutron Fluxes Required to Destroy MAs and non-Fissile Isotopes of Pu
Howell: The last 2 columns are very interesting, however during first reading I had no idea if the fluxes are "reasonable". With later reading: Section I.A p05h1.0 of the paper addresses this : the "set" neutron flux of 10^(+14) n/cm^2/s is typical for current reactors.
p10
p42h0.00 "Using the mass inventory data based on the lattice physics data, and the refueling rates based on core physics calculations, estimates were obtained for the net consumption rate of MAs in the 60 blanket channels and the net production rate of MAs in the 320 seed channels. ... If the production of other actinides (Howell: besides Am, Cm]) in the blanket channels are taken into account, such as Np, and also the plutonium, which has higher fractions of Pu-238 (up to 70 wt% for CC-MA-04), Pu-240 (up to 80 wt% for CC-MA-07) and Pu-242 (up to 19 wt% for CC-MA-03), then 90 to 121 channels of AmO 2 -based blanket fuel, and 18 channels of CmO 2 -based blanket fuel would be needed to achieve net zero production of MAs (if the Pu in the spent blanket bundles were treated as MAs as well). These estimates neglect the production of Pu in the seed fuel, which has a relatively low content of Pu-238 (~0.1 wt%) and Pu-242 (~1.5 wt%), and which could be recycled and consumed in (Pu,Th)O2 fuel bundles, as demonstrated in previous studies [4]."
Howell: ???
p42h0.25 "spent fuel is likely to be stored for at least 10 years before attempting to extract and recycle MAs"
Howell: ouch - what are secondary materials contamination issues like for this?
which processes? : [crush-grind-float, roast, [acid, base, oxidize, reduce] leach, volatilize [Cl, F], ion exchange, solvent extraction, gas diffusion, centrifuge, precipitate, filter, sinter, ???]
18 mixed oxide thorium based fuels Ashlea Colton, Blair Bromley Jul2018 "Lattice physics evaluation of mixed oxide thorium based fuels for use in pressure tube heavy water reactors" Annals of Nuclear Energy 117(2):259-276, https://www.researchgate.net/publication/326100919_Lattice_physics_evaluation_of_35-element_mixed_oxide_thorium-based_fuels_for_use_in_pressure_tube_heavy_water_reactors


ABSTRACT
A series of 2-D lattice physics depletion studies were carried out as part of conceptual scoping studies to evaluate thorium-based fuels in Heavy Water Reactors. Fuel bundles used for the study consisted of 35 fuel elements arranged in two outer rings comprised of thorium mixed with a fissile driver of either reactor grade plutonium (3.5-4.5 wt% PuO 2 ), low enriched uranium (5 wt% 235 U/U, 40-50 wt% LEUO 2 ) or uranium-233 (1.8 wt% 233 UO 2 ). Lattice physics- based estimated for fuel temperature coefficients and coolant void reactivity values were found to be lower than those for conventional natural uranium HWR fuel. The low burnup LEU/thorium fuel bundle shows the most promise as a feasible means of generating U-233 in a PT-HWR from a safety and fuel performance perspective.

Howell [question, comment]s :
- none for the moment. I am unfamiliar with many of the key [measurements, criteria] and they normally would be, so I would have to put in much more time to understand the significance of results. Sounds like one can operate current equipment with Thorium safely, and reasonable output?


Bromley, Alexander Dec2018 thorium and depleted uranium in sub-critical GC-PTR

Blair Bromley, Jude Alexander Dec2018 "Assessment of fast-spectrum blanket lattices for breeding fissile fuel from thorium and depleted uranium in an externally driven sub-critical gas-cooled pressure tube reactor" CNL Nuclear Review 7(2):201-220, DOI:10.12943/CNR.2018.00010 https://www.researchgate.net/publication/329346351_ASSESSMENT_OF_FAST-SPECTRUM_BLANKET_LATTICES_FOR_BREEDING_FISSILE_FUEL_FROM_THORIUM_AND_DEPLETED_URANIUM_IN_AN_EXTERNALLY_DRIVEN_SUB-CRITICAL_GAS-COOLED_PRESSURE_TUBE_REACTOR

ABSTRACT
To ensure long-term nuclear energy security, it is advantageous to consider the use of externally driven sub- critical systems for producing fissile fuel to supply fleets of thermal-spectrum reactors as an alternative to using fast- spectrum or thermal-spectrum breeder reactors.

Computational/analytical neutronics and heat transfer studies have been carried out for gas-cooled fuel bundle lattices with mixtures of fertile thorium and depleted uranium (DU) that could be used in the blanket region of a sub-critical fast reactor driven either by a fusion reactor in a hybrid fusion-fission reactor (HFFR) system, or an accelerator-based spallation neutron source in an accelerator driven system (ADS). The HFFR or ADS concept envisioned is one with a simple cylindrical geometry. The annular-cylindrical blanket is approximately 10 m long, can be made 2–5 m thick (1.0 m ≤ R blanket ≤ 3.0 m to 6.0 m), and is filled with a repeating square lattice of pressure tubes filled with 0.5 m long fuel bundles that are made of (DU,Th)O 2 , with various mixtures of Th and DU, and refuelled periodically online. Although using blankets made of pure DUO 2 or ThO 2 are viable options to analyze, mixing DUO 2 with ThO 2 can help alleviate any potential proliferation concerns, since any 233 U produced from breeding will be denatured by the presence of 238 U in (DU, Th)O 2 .

Lattice calculations demonstrate that the total fissile content in the fuel after an extended period of burnup (50 MWd/kg) will be approximately the same, regardless of the mixture of DU and thorium used, and that the content of americium and 232 U in the irradiated fuel will be <0.01 wt%/initial heavy metal.


Howell - same abstract as for Colton, Bromley Jul2018???


Colton, Bromley Mar2021 PT Heavy Water Reactor to Destroy Americium and Curium

Ashlea Colton, Blair Bromley Mar2021 "Reactor Physics Assesalternative lattice physics code and associated nuclear data set, such as SERPENT [61 ]sment of Potential Feasibility of Using Advanced, Nonconventional Fuels in a Pressure Tube Heavy Water Reactor to Destroy Americium and Curium" March 2021 Nuclear Technology 207(1):1-23, DOI:10.1080/00295450.2020.1853466, https://www.researchgate.net/publication/350481335_Reactor_Physics_Assessment_of_Potential_Feasibility_of_Using_Advanced_Nonconventional_Fuels_in_a_Pressure_Tube_Heavy_Water_Reactor_to_Destroy_Americium_and_Curium

Table I. Expected Time Scales and Thermal Neutron Fluxes Required to Destroy MAs and non-Fissile Isotopes of Pu
Actinide Half Life (years) Dominant Decay Mode Years to Decay naturally
to 1.0×10^-4 n/cm^2/s
*_ 		Thermal σ_absorb (barns) Thermal σ_fission (barns) Fission Spectrum σ_fission (barns) Years to Reduce to
1.0×10^-4 n/cm^2/s
with
1.0×10^+14 n/cm^2/s
Neutron Flux
**		 Flux to Reduce to
1.0×10^-4 n/cm^2/s
within 100 years
**
Np-236 154,000 EC to U-236*** 2,046,742 3073.9 2,453 1.92 1.0 9.50E+11
Np-237 2,144,000 Alpha to Pa-233 28,494,906 144 0.019 1.34 20.3 2.03E+13
Pu-238 87.7 Alpha to U-234 1,166 473.2 15.18 1.99 6.1 5.64e+12
Pu-240 6564 Alpha to U-236 87,239 263.7 0.053 1.36 11.1 1.11e+13
Pu-242 373,300 Alpha to U-238 4,961,357 16.8 0.0023 1.13 173.6 1.74e+14
Am-241 432 Alpha to Np-237 5,744 534.7 2.711 1.38 5.5 5.37e+12
Am-242m 141 IT to Am-242*** 1,874 7460. 6235 1.84 0.4 3.71e+11
Am-243 7,370 Alpha to Np-239 97,951 70.6 0.103 1.08 41.4 4.13e+13
Cm-243 29 Alpha to Pu-239 387 662.8 549.6 1.94 4.4 3.27e+12
Cm-244 18 Alpha to Pu-240 241 14.3 0.906 1.57 110.6 1.2e+14
Cm-245 8,500 Alpha to Pu-241 112,970 1961.4 1674 1.74 1.5 1.49e+12
Cm-246 4,760 Alpha to Pu-242 63,263 1.3 0.124 1.23 2212 2.29e+15
Cm-247 15,600,00 00 Alpha to Pu-243 207,332,337 121. 70.8 1.91 24.1 2.41e+13
Cm-248 348,000 Alpha to Pu-244 4,625,106 2.6 0.328 1.25 1,118 1.12E+15

*_ Calculated using equation N(t)/N(0) = exp(-1.0*0.693/(T_half*t))
** Calculated using equation N(t)/N(0) = exp(-1.0*0.693/(T_half*t) + σ*φ*t)).
*** Decay mode notes:
Fig 5 Net Change in Mass of U, Pu, and MAs in Seed and Blanket Bundle Concepts at Exit Burnup

Howell [question, comment]s :
Are [Am, Cm] already present in the feed fuel at very low levels, or are they produced during PT-HWR operation, either from the fuel or reactor materials of construction? The paper addresses the following issues after the introductory part :
What are the comparative costs associated with :
  • [extraction, concentration] of [Am, Cm] from spent fuel
  • running the transmutation of the [Am, Cm] target bundles in a PT-HWR reactor (perhaps for years to get acceptable levels)?
Does this require a dedicated PT-HWR, or can the [Am, Cm] bundles be placed with normal fuel in a normal CANDU over decades until threshold levels are obtained?
p02h0.45 Abstract "The potential to achieve net zero production of Am and Cm in a single thermal-spectrum reactor, such as a PT-HWR, could help eliminate the need to build and qualify a Deep Geological Repository (D
p03h0.40 Introduction "Long-term isolation and storage of MAs in a Deep Geological Repository (DGR) to ensure the protection of the environment for millions of years is considered scientifically and technologically feasible [1], [2]. However, it may be preferable to separate and destroy Am and Cm through direct fission, or transmutation into fissile isotopes by neutron capture, followed by fission."
Howell: Bernie Gorsky of NRCan/CANMET Mining in Ottawa (the divisions and branches have changed names a lot) did testing of core samples for DGRs. I was ignorant of the problems, but he took time to get me to realise that every time you create an openeing in rock formations, there is wall damage. Natural processes can also do that. As I understood it >10-20 years ago, the thinking was to store material in sturdy containers in DGRs, so that it could be removed perhaps decades later for reprocessing, or moving to a better site. Getting rid of it is nice if affordable - my occasional excursions into themes of history suggest that [political systems, societies] don't last forever either (normally, perhaps more like decades to a couple of centuries?).
p04h0.00 Table I Expected Time Scales and Thermal Neutron Fluxes Required to Destroy MAs and non-Fissile Isotopes of Pu
Howell: The last 2 columns are very interesting, however during first reading I had no idea if the fluxes are "reasonable". With later reading: Section I.A p05h1.0 of the paper addresses this : the "set" neutron flux of 10^(+14) n/cm^2/s is typical for current reactors.
p10
p42h0.00 "Using the mass inventory data based on the lattice physics data, and the refueling rates based on core physics calculations, estimates were obtained for the net consumption rate of MAs in the 60 blanket channels and the net production rate of MAs in the 320 seed channels. ... If the production of other actinides (Howell: besides Am, Cm]) in the blanket channels are taken into account, such as Np, and also the plutonium, which has higher fractions of Pu-238 (up to 70 wt% for CC-MA-04), Pu-240 (up to 80 wt% for CC-MA-07) and Pu-242 (up to 19 wt% for CC-MA-03), then 90 to 121 channels of AmO 2 -based blanket fuel, and 18 channels of CmO 2 -based blanket fuel would be needed to achieve net zero production of MAs (if the Pu in the spent blanket bundles were treated as MAs as well). These estimates neglect the production of Pu in the seed fuel, which has a relatively low content of Pu-238 (~0.1 wt%) and Pu-242 (~1.5 wt%), and which could be recycled and consumed in (Pu,Th)O2 fuel bundles, as demonstrated in previous studies [4]."
Howell: ???
p42h0.25 "spent fuel is likely to be stored for at least 10 years before attempting to extract and recycle MAs"
Howell: ouch - what are secondary materials contamination issues like for this?
which processes? : [crush-grind-float, roast, [acid, base, oxidize, reduce] leach, volatilize [Cl, F], ion exchange, solvent extraction, gas diffusion, centrifuge, precipitate, filter, sinter, ???]

Wojtaszek, Bromley Jul2022 Uranium-Based Oxy-Carbide in Compact HTGC Reactors

Daniel Wojtaszek, Blair Bromley Jul2022 "Physics Evaluation of Alternative Uranium-Based Oxy-Carbide Annular Fuel Concepts for Potential Use in Compact High-Temperature Gas-Cooled Reactors" Journal of Nuclear Engineering and Radiation Science 9(1), DOI:10.1115/1.4055009 https://www.researchgate.net/publication/362027621_Physics_Evaluation_of_Alternative_Uranium-based_Oxy-Carbide_Annular_Fuel_Concepts_for_Potential_Use_in_Compact_High-Temperature_Gas-Cooled_Reactors



Wojtaszek, Bromley Aug2023 Plutonium-Thorium Fuels with 7LiH Moderator

Daniel Wojtaszek, Blair Bromley Aug2023 "Reactor Physics Assessment of Annular Plutonium-Thorium Fuels for Use in Prismatic Fuel Blocks in a HTGR-SMR with a Hydrogen-Based Moderator (7LiH)" 2023 Mathematics and Computations Conference (MC2023) (CNS and ANS), Niagara Falls, Ontario, Canada. Authors: Canadian Nuclear Laboratories (since 2014; formerly AECL Chalk River Laboratories 1952-2014) https://www.researchgate.net/publication/373257928_DRAFT_MC2023_Conference_Proceedings_Reactor_Physics_Assessment_of_Annular_Plutonium-Thorium_Fuels_for_Use_in_Prismatic_Fuel_Blocks_in_a_HTGR-SMR_with_a_Hydrogen-Based_Moderator_7LiH