Isotope Selection:

1
Isotope Selection
Lunar Radioisotope Power System
Presented by Caleb Robison, Dan
Osterberg Contributions from Jeff Katalenich,
Joel Sasser, Palani Ramu
2
Outline
1.1 Mission Requirements and Isotope
Consideration Factors 1.2 Isotope Selection
Process 1.3 Initial Candidates 1.4 Final
Candidates 1.4.1 Power Densities 1.4.2 Decay
Modes and Shielding 1.4.3 Production/Availability
1.5 Curium Considerations 1.5.1 Radiation
Concerns 1.6 Sr-90, Pu-238, Cm-244 Comparison 1.7
Conclusions
3
Mission Requirements and Isotope Consideration
Factors
Minimum of 5 year mission Single RPS to
provide 2.5 kWe until EOM Realistic isotope
extraction/production costs Heat source
assembly must survive re-entry and impact
Minimal dose to workers during fabrication
Total mass (including shielding) reasonable for
launch into space
Isotope Properties Considered Half Life
Power Density (including compound form and
isotopic concentration) Dose rate / Shielding
requirement Production / Availability / Cost
4
1.2 Isotope Preliminary Selection
5
1.3 Initial Candidates
Power Density
Availability
Power Density
Notes 1) Power Density of Compound at 5 years
(isotopic considered for Sr-90, Ru-106, Ce-144,
Pu-238, Cm-244) 2) Spent fuel concentrations at
5 years out of reactor
6
Availability
Isotopic Separation
Availability
Availability
Notes 1) Power Density of Compound at 5 years
(isotopic considered for Sr-90, Ru-106, Ce-144,
Pu-238, Cm-244) 2) Spent fuel concentrations at
5 years out of reactor
7
Availability
Notes 1) Power Density of Compound at 5 years
(isotopic considered for Sr-90, Ru-106, Ce-144,
Pu-238, Cm-244) 2) Spent fuel concentrations at
5 years out of reactor
8
1.4 Final Candidates
On the basis of power density at EOM, radiation
dose, and availability, the most suitable
isotopes for use in a Lunar RPS
9
1.4.1 Power Densities
In order to maintain 2.5 kWe in a single RTG unit
for a 5 year mission, a volumetric power density
of 8.66 W/cc at EOM is required.
Note Values based on assumption of 25 heat to
electricity conversion
10
4.1.2 Decay Modes and Shielding
11
4.1.3 Decay Modes and Shielding
12
4.1.4 Production and Availability
Pu-238
Purchase - most cost effective when
available in stockpile - must be produced if
widespread use continues into the future
Breed from Np-237 - approximately 400kg
available in storage at the INL - ATR could be
used for Pu-238 production - cost of 150-200 M
for infrastructure and 15-20 M/year
operational - rate of approximately 5kg/yr
Breed from Am-241 - Am-241 neutron capture ?
Am-242 ? Cm-242 decays by alpha ? Pu-238 -
Spent fuel has 60 - 1370g/MTIHM of Am-241
Source Stephen Johnson, INL
13
4.1.4 Production and Availability
Sr-90
Stockpile - widely available in United
States stockpiles - 2.3 million Ci as of 2007 -
exists in the form of SrF2 - separated from
other waste - Sr-90 composes 50-60 of Strontium
isotopically - stored at Hanford Site
14
4.1.4 Production and Availability
Cm-244
Recover from vitrified storage at Savannah
River Site - F Canyon Tank at SRS contained
15,000 L fission products in dilute nitric acid
was vitrified into a lanthanide borosilicate
glass - 99 recovery of Cm by means of 15.7M
nitric acid - approximately 2.56 kg of Cm-244
available from this source Extraction from
spent fuel - 27.2 - 85.6 g/MTIHM in UO2 fuels -
time dependent - Much higher amounts in MOX or
plutonium fuels - Urex4 process has the
ability to separate Cm - Depends on change in
U.S. fuel cycle strategy - 1500 ton/year
facility could yield 30 40 kg/year
15
4.1.4 Production and Availability
U-232
Neutron irradiation of U-233 - Yields gram
quantities per year too expensive Proton
beam Th-232 - Yields gram quantities per year
too expensive Extract from spent fuel -
Small amounts low isotopic percent Neutron
irradiation from Th-230 - Small yield
moderate quantities available
16
1.5 Curium Considerations
Neutron Flux - 4.3 x 107 neutrons/second/gram
Fuel Form - Foil - Cm2O3 - Cm Hydride
Helium Gas - 1 gm Cm-244 evolves 45.9cm3 He gas
in 1 half life (18.1 yr) Reprocessing
Concerns - Cm-244 generates much more heat in
concentrated solutions than has been dealt with
in wide scale production (up to 700Wt/L) -
Concentrated streams of curium in reprocessing
may approach critical mass of Cm-245

17
1.5.1 Radiation Concerns
Cm-244
Major Radiations 5.80 MeV a (76.4)
5.76 MeV a (23.6) 42.8 keV ? (2.6E-4)
4.3 x 107 n/sec/gm
Our shielding calculations indicate that 10cm of
borated polyethylene will supply sufficient
neutron shielding.
18
1.5.1 Radiation Concerns

• Pu-238 (PuO2) modeled for baseline and accuracy
  (compared to existing papers)
• Neutron - 0.3610.002 mrem/hr
• Photon - .650.03 µrem/hr
• Total .36368 mrem/hr
• Cu-244 (Cu2O3)
• Neutron 2.010.03 rem/hr
• Photon - 56.50.2 mrem/hr
• Total 2.0967 rem/hr

Above GPHS Model (MCNP VisEd) Left Dose vs.
Mass of Pure LiH Shielding GPHS General
Purpose Heat Source
19
1.6 Sr-90, Pu-238, Cm-244 Comparison
20
1.7 Conclusions
Sr-90 and Pu-238 would work for this
application - Power density is low - Launch
masses will be high - Sr-90 is the cheapest
isotope to acquire Cm-244 is the best
choice - Power density is high - Launch
masses will be low - Can be recovered through
proven methods in large quantities - Sources
exist and will source concentration will increase
with time U-232 would be ideal for this
application - Long string of alpha decay
U-232 ? Th-228 ? Ra-224 ? Rn-220 ? Po-216 ?
Pb-212 - Low launch mass - Production
limited
21
Questions?
22
Evaluation of General Purpose Heat Source Powered
Stirling Technologies for a2.5 kWe Lunar Surface
Power SourcePresented By Chris Miller and Troy
ReissContributions by Chris Miller, Troy
Reiss, Jeff Katalenich, Logan Sailer, Caleb
Robison August 10, 2007
23
Contents

• Overview of existing space power conversion
  technology
• Justification for selection of Stirling power
  conversion system
• Review of current Stirling systems for space
  applications
• Concepts offering potential improvement to
  existing technology

24
Radioisotope Thermoelectric Generator (RTG)

• Flown on Galileo, Ulysses, Cassini, and New
  Horizons missions
• Powered by Pu-238 General Purpose Heat Source
  (GPHS)
• Produced approximately 300 We using 572 silicon
  germanium thermoelectric elements
• Thermal power 4400-4500 W from 18 GPHS modules,
  mass 55.9 kg, specific power 5.5 We/kg,
  efficiency 6.7-6.8
• Reliable, but electrical power output well below
  lunar surface mission requirement

25
Stirling Power Conversion System

• Other power conversion systems explored to
  provide greater power output than RTG
• Stirling and Brayton systems found to be capable
  of supplying desired power
• Stirling determined to be best power conversion
  system due to better scaling at desired power
  level
• Better scaling results in lower mass and higher
  efficiency than Brayton cycle based power
  conversion system

26
Stirling vs Brayton1 kW to 10 MW
27
Existing TechnologyDual-Opposed Technology
Demonstration Converters

• Hot end temp 923K
• Cold end temp 333K
• 2 55 We Stirling
• Led to development of the SRG110
• Achieved 26,000 hours of operation
• Specific power of 3.5 We/kg

28
Sunpower Advanced Stirling Converter (ASC)

• Resulted from the SRG110
• Free piston design
• Weight and size reduction
• 88 We at 38 Efficiency
• Hot end 1123K

29
Advanced Stirling Radioisotope Generator(ASRG)

• Lifetime of 14 years plus 3 years of storage
• BOM Power Output 140 We
• EOM 126 We (14 yrs)
• Projected Mass 20.24 kg
• Projected Specific Power 7.0 We/kg (Using Pu)
• 72.5 cm L x 41 cm H x 29.3 cm W
• Beryllium housing
• Future model projects 8.5 We/kg

30
Stirling Lunar Power System (LPS)

• Stacking GPHS modules has limits as distance
  increases from the Stirling converter
• Radial configuration allows all GPHS modules to
  be in contact with the hot shoe
• Can accommodate many GPHS modules
• Ni-200 hot shoe for high thermal conductivity

31
Potential Improvements

• Problems
• Current technology fails to meet the 2.5 kWe goal
• Multiple units of the ASRG to meet power
  requirements would be too heavy
• Current heat rejection systems subjected to lunar
  day/night cycles
• Changing environmental conditions changes cold
  end temperature, affects Stirling performance
• Solution
• Development of Stirling converters with a higher
  electrical power output
• Designs to incorporate multiple Stirling
  converters were developed
• Concept to reduce cold end variation during
  changing environmental conditions proposed

32
Development of New Stirling Converters

• Nasa has recently funded development of a 5 kWe
  free-piston stirling converter for lunar
  application
• Sunpowers EG1000 is a 1.2 kWe free piston
  stirling converter has been in use for DOD
  applications for several years
• Infinia Corp. is working to develop a 3 kW
  free-piston stirling convertor for solar
  applications

33
Tri-core with Two Stirling Converters

• Based on LPS configuration
• Used to determine capability of other
  configurations
• Results would be baseline for other designs

34
Quad-core and Octagon-core

• Based on LPS design
• Ability to attach multiple Stirling converters
• Stirling converters with higher power output
  could be attached

35
Heat Pipe Concept

• 4 converters .75 kWe each
• 5 GPHS blocks (Cm 244)
• Heat pipes couple GPHS modules to Stirling hot
  ends
• Working fluid transport via capillary forces in a
  wicking structure
• Heat rejected to surroundings via cold end
  radiators
• Possible power output of 2.5 to 3 kWe
• Designed to be easily assembled prior to or after
  launch

36
Stirling Cooling

• All designs discussed thus far cooled from a cold
  flange attached to the cold end of the engines
• The cold flanges are coupled to the outer shell
  which acts as the radiator for the unit
• Issues arise from changing lunar day/night
  temperatures and lunar dust collection on
  radiators
• Method of heat rejection from cold end identified
  as major potential area for improvement of
  existing concepts

37
Sub-lunar Surface Heat Sink Concept

• Current lunar Stirling concepts exposed to
  changing environmental conditions during lunar
  day/night cycle
• Potential to eliminate this complication through
  use of constant -30 C sub-lunar surface
  temperature as heat sink
• Liquid metal or sulfur injected into bedrock or
  regolith during drilling operation
• Liquid diffuses into lunar material, providing
  higher thermal conductivity sink than lunar
  material alone
• Stirling cold end coupled to sink with high
  thermal conductivity material prior to freezing
• Heat rejected via conductive path to sink instead
  of radiator

38
Sub-lunar Surface Heat Sink Concept

• Advantages
• Elimination of fluctuating cold end temperature
  and power output
• No exposure of radiators to lunar dust
• Reduced shielding and insulation mass if entire
  assembly placed below lunar surface
• Potential mass savings from removal of radiators
• Remaining issues
• Thermal analysis must determine necessary size of
  heat sink
• Cold end temperature must be determined and
  compared with cold end temperatures of current
  concepts
• Tradeoffs between sulfur and metal sink must be
  determined and evaluated
• Physical location of sink, converter, and GPHS
  units must be determined

39
Conclusions and Future Work

• Stirling converters best power conversion option
  for 2.5 kWe lunar surface radioisotope power
  system
• Free piston Stirling should be basis for such
  systems
• Great potential for improvement of existing
  Stirling systems through utilization of sub-lunar
  surface heat sink
• Extensive modeling and thermal analysis must be
  performed on all proposed concepts to determine
  if they offer improvements over existing Stirling
  systems

40
Questions?
41
RPS Cooling Options on the Moon

• Holly Szumila,
• Mookesh Dhanasar
• Benjamin Schreib
• Center for Space Nuclear Research
• August 10, 2007

42
Outline

• Assembly / In-transit cooling (active)
• Cooling options on the moon
• Surface
• Sub-surface
• Analysis
• Thermal models (1-D analytical, 2-D numerical)
• Compare different sources
• Conclusions
• Questions

43
RPS Cooling (Assembly and In-Transit)

• Already have assembly active cooling.
• In transit cooling systems already exist.
• Include fins.

44
Cooling options on the moon

• Lunar Surface
• Lunar Sub-surface
• Regolith
• Bedrock

45
Lunar Surface

• Lunar Surface
• Unstable temperatures
• Micrometeorites
• Radiation
• Lunar Dust

46

• Radiation considerations
• Solar wind, peak solar flares, galactic cosmic
  radiation.
• Primary concern solar cosmic radiation, or solar
  flares.
• Heavy ion fluxes not accounted for using MCNPX,
  but can travel through layers of shielding and
  spallation effects (high energy neutron fluxes).
• Proton and neutron fluxes can only cause heat
  deposition to RTG on nano-Watt magnitude.
• Lunar Dust
• Micrometeorites
• Caused thin films on Apollo structures, thought
  to pile a great deal and cause wear to metal over
  extended periods of time.

47
Lunar Surface - Conditions
LUNAR AND MARTIAN ENVIRONMENTAL INTERACTIONS WITH
NUCLEAR POWER SYSTEM RADIATORS Maria E.
Perez-Davis and James R. Gaier NASA Lewis
Research Center, Cleveland, OH 44135 Cynthia M.
Katzan Sverdrup Technology, Inc., Lewis Research
Center Group
48
Lunar Sub-Surface (Regolith)

• Lunar sub-surface (Regolith)
• Thermal shield.
• Constant temperature.
• Radiation shield.
• Shielding against micro-meteorites.

49

• Lunar Bedrock
• Difficulty in drilling
• Less known on bedrock
• Constant temperature sink

50
Lunar Sub-Surface Thermal analysis (Regolith)
51
RPS Thermal Analysis-Model
52
RPS Thermal Analysis-Assumptions

• The analysis is carried out in 1-D only.
• Steady-State conditions apply.
• Heat generation is constant and uniform.
• The thermal conductivity for the material is
  constant.
• The RTG is symmetric about its centerline.
• There is a conduction-convection interface with
  the outer surface of the RTG and the medium.

53
RPS Thermal Analysis Theory (Analytical)
54
RPS 1-D Temperature Profile
55
Surroundings Analysis - Model
56
Surroundings Analysis - Assumptions

• 1-D heat transfer.
• Steady State conditions apply.
• There is no heat generation in the region of
  interest.
• There is no bulk fluid motion, so heat transfer
  is a special case of conduction.
• The temperature at the surface of the RTG is
  known.
• The ambient temperature is known.

57
Surroundings Analysis Theory (Analytical)
58
RPS Surrounding Area Temperature Profile
59
Surroundings 2-D Analysis (Numerical)
60
Surroundings 2-D Analysis (Point Source) Results
61
Surroundings 2-D Analysis (Numerical) Results
62
Compare Various RPS Sources
63
Compare Various RPS Sources
64
Compare Various RPS Sources
65
Compare Various RPS Sources
66
Conclusion

• From our research it is desirable to have the RPS
  buried in the regolith.
• From the thermal analysis, a simple heat transfer
  tool was created.
• It was used to determine the thermal profile for
  a variety of sources.
• From this analysis it is observed that for the
  commonly used isotope source, approximately 0.8 m
  of sulfur is required before phase change occurs.

67
Future Work

• Refine model.
• Detail 3-D
• Investigate various convective mediums.

68
Questions
69
Acknowledgments

• Dr. Steve Howe
• John Bess, Jon Webb
• Ms. Kristi Bailey
• INL
• 2007 CSNR Summer Fellows

70
Isotope Selection
Radioisotope Powered Vehicle
Presented by Jeff Katalenich Contributions
from Joel Sasser, Caleb Robison, Dan
Osterberg, Troy Reiss, Jeff Perkins
71
Outline
1.1 Mission Requirements and Isotope
Consideration Factors 1.2 Isotope Selection
Process 1.3 Initial Candidates 1.4 Final
Candidates 1.4.1 Power Densities 1.4.2 Decay
Modes and Shielding 1.4.3 Production/Availability
1.5 Isotopic Mass Requirements 1.6 Ru-106,
Ce-144, Po-210, Cm-242 Considerations 1.6.1 Reali
stic Power Densities 1.6.2 Power Density
Decay 1.7 Curium Considerations 1.7.1 Radiation
Concerns 1.7.2 Reprocessing Options 1.8
Conclusions
72
1.1 Mission Requirements
Minimum flight time of 7-12 months Isotope
shall provide 20-40 kWt Realistic isotope
extraction/production costs Heat source
assembly must survive re-entry and impact
Minimal dose to workers during fabrication
Total mass (including shielding) reasonable for
our flight unit and for launch into space
Isotope Properties Considered Half Life
Power Density (including compound form and
isotopic concentration) Dose rate / Shielding
requirement Production / Availability / Cost
73
1.2 Isotope Preliminary Selection
74
1.3 Initial Candidates
Availability
Gamma Dose
Notes 1) Power Density of Compound at 5 years
(isotopic considered for Sr-90, Ru-106, Ce-144,
Pu-238, Cm-244) 2) Spent fuel concentrations at
5 years out of reactor
75
Gamma Dose
Isotopic Separation
Notes 1) Power Density of Compound at 5 years
(isotopic considered for Sr-90, Ru-106, Ce-144,
Pu-238, Cm-244) 2) Spent fuel concentrations at
5 years out of reactor
76
Isotopic Separation
Availability
Availability
Notes 1) Power Density of Compound at 5 years
(isotopic considered for Sr-90, Ru-106, Ce-144,
Pu-238, Cm-244) 2) Spent fuel concentrations at
5 years out of reactor
77
Availability
Availability
Notes 1) Power Density of Compound at 5 years
(isotopic considered for Sr-90, Ru-106, Ce-144,
Pu-238, Cm-244) 2) Spent fuel concentrations at
5 years out of reactor
78
Availability
Notes 1) Power Density of Compound at 5 years
(isotopic considered for Sr-90, Ru-106, Ce-144,
Pu-238, Cm-244) 2) Spent fuel concentrations at
5 years out of reactor
79
1.4 Final Candidates
On the basis of power density at EOM, radiation
dose, and availability, the most suitable
isotopes for use in the RPV are
80
1.4.1 Power Densities
Analysis showed that a minimum power density of
0.40 W/g is necessary based on mass requirements.
Note Compound Power Density Values Account for
Isotopic Concentration and are taken at 1.5 years
81
1.4.2 Decay Modes and Shielding
82
1.4.3 Production and Availability
Sr-90
Stockpile - Widely available in United
States stockpiles - 2.3 million Ci as of 2007 -
Exists in the form of SrF2 - Separated from
other waste - Sr-90 composes 50-60 of Strontium
isotopically - Stored at Hanford Site
83
1.4.3 Production and Availability
Ru-106
Extraction for Spent Fuel - 2.086kg/MTU
Ruthenium in UO2 spent fuel - Ru-106 has 7
isotopic concentration - Possibility of minor
isotopic separation?
84
1.4.3 Production and Availability
Ce-144
Stockpile - 2.28kg/MTU Cerium in spent UO2
fuel - Ce-144 has isotopic concentration of
9-13 - Possibility of minor isotopic
separation?
85
1.4.3 Production and Availability
Po-210
Neutron Irradiation of Bi-209 - Bi-209 gtgt
Bi210 gtgt 5 day half life gtgt Po-210 - Bi-209
composes 100 of natural Bismuth - Bi-209 has
capture cross section of 0.3 microbarn - Ni-210
and Po-210 have capture cross sections in the
millibarns - Large batch production - High
neutron flux required
86
1.4.3 Production and Availability
Pu-238
Purchase - Most cost effective when
available in stockpile - Must be produced if
widespread use continues into the future
Breed from Np-237 - Approximately 400kg
available in storage at the INL - ATR could be
used for Pu-238 production - Cost of 150-200 M
for infrastructure and 15-20 M/year
operational - Rate of approximately 5kg/yr
Breed from Am-241 - Am-241 neutron captures ?
Am-242 which decays ? Cm-242 which decays ?
Pu-238
Source Stephen Johnson, INL
87
1.4.3 Production and Availability
Cm-242
Neutron irradiation from Am-241 - Am-241
gtgt Am-242 gtgt 16 hour half life gtgt Cm-242 -
Am-241 available in 78-87 pure isotopic form out
of spent fuel aged 10-30 years in quantity of
929-1560 g/MTIHM - Best yield at neutron flux of
7 x 1014 n/sec/cm2 similar to ATR - Period of
4-6 months gives best yield - Cm-242 is
available in spent fuel, but in small isotopic
percentages - Neutron irradiation yields purest
form of Cm-242
88
1.4.3 Production and Availability
Cm-244
Recover from vitrified storage at Savannah
River Site - F Canyon Tank at SRS contained
15,000 L fission products in dilute nitric acid
was vitrified into a lanthanide borosilicate
glass - 99 recovery of Cm by means of 15.7M
nitric acid - approximately 2.56 kg of Cm-244
available from this source Extraction from
spent fuel - 20-30 g/MTU in UO2 fuels -
400-1000 g/MTU in Pu recycled fuels -
reprocessing only UO2 fuels would allow for
production of 15kg/yr - reprocessing 25 MOX and
75 UO2 allows production of 200kg/yr - Urex
process separates Am/Cm from waste stream -
primary concerns include thermal loading and
neutron dose
89
1.5 Isotopic Mass Requirements
Note Assumes a necessary thermal output of 20 kW
at EOM
90
1.6 Ru, Ce, Po, Cm-242 Considerations
Ru-106, Ce-144, Po-210, and Cm-242 appear to be
excellent isotopes for space missions, but the
following variables need to be taken into
consideration Mission Timeline
Achievable Power Densities Power Densities
Changes with Time
91
1.6.1 Realistic Power Densities
When calculating power densities it is necessary
to take isotopic concentration and compound form
into account in addition to energy per decay.
Note Power densities taken at 1 month after
fabrication
92
1.6.2 Power Density Decay
The short half lives of Ru-106, Ce-144, Po-210,
and Cm-242 make them preferable for short term
missions
93
1.7 Curium Considerations
Neutron Flux - 4.3 x 107 neutrons/second/gram
- Effectively shielded by Cd Availability -
Cost for recovery facility - Access to fuels
with high Cm content - Reprocessing prospects in
the US Fuel Form - Foil - Cm2O3 - Cm
Hydride Helium Gas - 1 gm Cm-244 evolves
45.9cm3 He gas in 1 half life (18.1 yr)
Critical Mass - Concentrated streams of curium
in reprocessing may approach critical mass of
Cm-245
94
1.7.1 Radiation Concerns
Cm-244
Major Radiations 5.80 MeV a (76.4)
5.76 MeV a (23.6) 42.8 keV ? (2.6E-4)
4.3 x 107 n/sec/gm
Cm-244 generates much more heat in concentrated
solutions than has been dealt with in wide scale
production (up to 700Wt/L)
Our shielding calculations indicate that 10cm of
borated polyethylene will supply sufficient
neutron shielding.
95
1.7.1 Radiation Concerns MCNPX

• 16.5 kg Cm2O3 required for 40 kW (thermal)
• Unshielded dose 8.6 rem/hr
• 300 kg LiH 50 mrem/hr
• 90 kg LiH and 190 kg of a Gd/U mixture 60
  mrem/hr
• Pure LiH may be more attractive than including
  Gd/U if volume is not important 300 kg of LiH
  should fit inside the launch vehicle

Above RPV source model (MCNP VisEd)
dose cylinder and isotope shielding Left Dose
vs. Mass of pure LiH shielding
96
1.7.2 Reprocessing Options
Production Facilities - Studies in 1970s
suggested Barnwell Nuclear Fuel Plant as Cm
reprocessing site - Separate study suggested
modifications to Hanford, Savannah River, Idaho
Chemical Processing Plant, and Oak Ridge -
These studies were performed before 1979 and made
the assumption that Pu recycled fuel would
become widespread in the US Recovery -
Recover 90 of Cm in waste - Concentration in
spent fuel most dependent on Pu-242
concentration - Best to process fuel 6-18 months
out of the reactor, but up to 4 years
97
1.8 Conclusions
Sr-90 and Pu-238 could work for this
application - power density is low - launch
masses will be high - Sr-90 is the cheapest
isotope to acquire Cm-244 is the best
choice - power density is high - launch
masses will be low - can be recovered through
proven methods in large quantities - sources
exist and will source concentration will increase
with time Several issues arise with using
Ru-106, Ce-144, Po-210, or Cm-242 -
Fabrication of an isotope that is very hot in
concentrated form - Mission timeline
critical - Higher shielding requirements for
Ru-106 and Ce-144
98
Questions