activation of air and concrete in medical isotope ... · “the measurement of neutron energy...
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Activation of Air and Concrete in Medical Isotope Production Cyclotron Facilities
Adam Dodd Senior Project Officer
Accelerators and Class II Prescribed Equipment Division (613) 993-7930 or [email protected]
CRPA 2016, Toronto
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Outline
• Motivation • Methods • Air activation • Concrete activation • Conclusions
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Motivation
• Neutrons produced in cyclotron target via (p,Xn) – can activate materials in vault – air activation can pose a radiological hazard to workers – concrete activation can become a disposal problem
• Why the renewed CNSC interest? – cyclotrons have become much more powerful – old days ~ 60 µA and 2 hour runs for F-18 FDG production – now could have 500 µA and 6 hour runs for Tc-99m production – 25 times more activation!
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Methods (1/2)
Monte Carlo simulations – MCNP5 (1% statistics) – proton beams 18 & 24 MeV – simplified vault geometry – F-18 and Tc-99m neutron source spectra – assumed
isotropic – explore sensitivity of results to
• changes in vault design • source energy • source location • polyethylene (with/without boron) shielding around
targets
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Methods (2/2)
Collaboration with UOIT co-op students
– Rob Shackelton (air activation) – Devon Carr (concrete activation) – Audrie Ismail (concrete activation)
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Air activation
• Nitrogen 78%of air – N-14(n,p)C-14 (78% of air) – C-14: 5,730 year half-life, soft β – little radiological consequence
• Oxygen 21% of air – O-16(n,p)N-16 (21% of air, 10 MeV threshold) – N-16: 7s half life, 6 MeV γ – little radiological consequence
• Argon ~ 1% of air – Ar-40(n,γ)Ar-41 (0.93% of air) – Ar-41: 1.8h half life, hard β, 1.3 MeV γ
Dominant hazard is Ar-41 activity
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Geometry and Materials Go from complex to extremely simple Reveals essential features and save computing time
Iron
Concrete
Air
Void
x
y
Neutron source
12 m
6 m
2 m
3 m
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Neutron Source Spectra (1/2)
• Neutron point source emulates target during irradiation (isotropic) • F-18 thick-target spectrum from Mendez et. al. [1]
– approximated by Maxwell fission spectrum – 150 logarithmically spaced energy bins
• Tc-99m spectrum from nested neutron spectrometer data [2] – Histogram representation
• All spectra automatically normalized by MCNP
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Neutron Source Spectra (2/2)
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Neutron Source Placement
• 3m x 5m x 5m vault
• Eight source positions
• Starting at origin, moving
diagonally into corner
• Considering statistical error, effect of source location is negligible
Effect of Source Placement on Ar-41 Production
-1.50%
-1.00%
-0.50%
0.00%
0.50%
1.00%
0 0.5 1 1.5 2 2.5 3 3.5
Distance from Origin (m)Pe
rcen
t diff
eren
ce fr
om m
ean
Ar-
41
prod
uctio
n
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Vault Size Linearity result of large neutron mean free path in air ~62 m
Deviation from linearity for irregular shapes and very small vaults
Ar-40 captures vs. cube bunker edge length
0 1 2 3 4 5 6 7 8
Cube edge length (m)
Ar-
40 C
aptu
re d
ensi
ty
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Cyclotron and Source Energy
• Iron cyclotron added to centre of the room – Slight decrease in production – Cyclotron density has no effect
• Source energy tests: (3m x 6m x 12m vault with cyclotron)
Neutron energy Ar-41 production/incident neutron/cm3
F-18 spectrum 3.0 E-5 Tc-99 spectrum 3.6 E-5 Isotropic 1 keV 5.2 E-5 Isotropic 0.025 eV 12.0 E-5
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Target Clamshell
• Polyethylene target shielding
– 5, 10, 50cm thickness – No boron
• Critical thickness
– Thin shield thermalizes neutrons – >10cm necessary to capture
• γ shielding requirements
Iron
Concrete
Air
Void
x
y
Source
12 m
6 m
2 m
3 m Polyethylene
clamshell
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Partition Walls
• Wall placement below resulted in 15% decrease in Ar-41 production
Iron
Concrete
Air
Void
x
y
Source
20 m
7 m
2 m
3 m
• Objects in vault reduce air activation
• Justifies simplified geometry
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Ar-41 Activity
• Results are in captures/neutron - How many neutrons per μA of beam? – IAEA TRS-468x gives saturation activity of F-18 at different beam
energies; at saturation • Neutron production rate = F-18 decays/s (Bq)
– extrapolated to 24 MeV → 1.6 E10 n/s/μA • For Mo-99 (NNS @19 MeV) → 3.2 E10 n/s/μA
• At saturation • F-18 at 150 μA - (3 x 5 x 5 m vault) ~2 mCi • Tc-99 at 750 μA – (3 x 6 x 12 m vault) ~ 12 mCi
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Dosimetry – Dose rates
All dose rates in μSv/h
F-18 @ 150 μA 3m x 5m x 5m
Tc-99 @ 750 μA 3m x 6m x 12m
External gamma3 5.1 17.3
Inhalation4 0.1 0.3
Skin5 0.9 2.2
• Assumes saturation production • Skin dose – upper limit – assumes no clothing
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Air Activation - Summary
1. Objects in vault reduces air activation • Justifies simplified geometry
2. Air activation is not a problem for F-18 • Results are for saturation of Ar-41 (half-life 1.8 h) • Runs are ~ 3 hours and typical F-18 runs are shorter
3. For Tc-99 may be a problem • Runs are ~ 6 hours and beam current ~ 3 times higher
4. Ventilation reduces problem dramatically • Less time to build up Ar-41 in vault and less exposure time –
1 hour air exchange time reduces dose by a factor of ~10
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Concrete Activation
This produces radioactive waste, affecting decommissioning costs
1. How deep does it go? 2. What do polyethylene layers (with/without
boron) do? 3. Is it on all inner vault surfaces? Or is it
localized?
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Literature Review (1/2) Decommissioning a cyclotron [6]
20-year-old 17 MeV Scanditronix cyclotron (~40 µA) • 40 tons of low-level radioactive waste including the concrete vault wall
• Activities with τ½ > 1 year Isotope Measured activity (Bq/g) UCL (Bq/g)
Co-60 0.068 0.1
Cs-134 0.005 0.1
Eu-152 0.083 0.1
Eu-154 0.010 0.1
Mn-54 * 0.016 0.1
Total 0.18 0.1 * Made by fast neutrons via (n,p) reaction rather than (n,γ)
UCL: unconditional clearance level at which material can be thrown out as non-radioactive,
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Literature Review (2/2)
Reactor study [7] Ordinary concrete sample in TRIGA reactor in Slovenia 30 minute exposure @ neutron flux of 6.8 1012 n/s/cm2
Principal activities found Eu-152 and Co-60 at 6 Bq/g Conclusion – concrete activation could be a
problem with the new cyclotrons
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Absorption depth
1 3 5 15 25 35 45 55 65 80 100
Neu
tron
cap
ture
den
sity
Concrete depth (cm)
Neutron capture density in concrete (F-18 production)
No Polyethylene
10 cm thick Polyethylene
10 cm thick BoratedPolyethylene
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Polyethylene around target Poly layer thickness (cm) Percentage of neutrons
captured in poly layer Percentage of neutrons captured in borated poly layer
5 17.1 32.0 10 52.0 65.8 15 72.5 82.9 20 82.6 91.1
• Results show how deep a sacrificial layer should be (if used) • Regular poly is almost as good as borated poly • Either option much cheaper than sacrificial layers in the vault • And can be put in after vault construction
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Spatial Distribution in Vault
We now know activation goes down to depth ~20cm But this was averaged over whole inner vault surface Is it on all inner vault surfaces ? Or is it localized? Investigated 1. Tc-99 vs F-18 2. Moving the source position inside the vault 3. Lateral distribution of neutron capture density within
1 side of the wall
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1. Neutron Escape Percentage for Regular Poly Around Target - Tc-99 versus F-18 source
1. F-18 neutrons escape the polyethylene layer more easily than Tc-99’s higher energy (more penetration)
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2. Moving the Source Position on Y-Axis With 20-cm Thick Polyethylene Layer
75 250 425 600cm
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2. Relative Capture Density of the Left Wall With Respect to Tc-99 Source
Position
y = 3.4e-0.002x R² = 0.99
0.0
0.5
1.0
1.5
2.0
2.5
3.0
3.5
0 100 200 300 400 500 600 700
Capt
ure
Den
sity
rel
ativ
e to
whe
n th
e so
urce
is in
m
iddl
e po
siti
on
Distance between source and left wall (cm)
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3. Lateral Distribution of Neutron Capture Density in Near Wall
0
50
100
150
200
250
-50 0 50 100 150 200 250 300
Cap
ture
den
sity
in th
e fr
ont w
all
Lateral distance on wall from target (cm)
Inverse Square Law
50cm away WO poly
50cm away WITH poly
middle WO poly
At 50-cm distance radius of activation ~ 150 cm (10%)
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Regulatory Impact
1) Compare cyclotron neutron flux with TRIGA reactor flux • After 1 year at full operation → 0.35 Bq/g of Eu-152 and
0.33 Bq/g for Co-60 (measurable) • After 25 years operation → 7 Bq/g for Eu-152 and
3 Bq/g of Co-60 2)100 x regulatory limit for disposal as non-radioactive waste 3) If no steps are taken, it will impact decommissioning cost and possibly financial guarantee
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Conclusions • Air activation – Not a big deal and easily controlled through
– ventilation – restricted access for a few hours (normal to allow cyclotron to cool off) – detection of Ar-41 by area monitor in vault
• Concrete activation – could be a challenge for decommissioning – localized to concrete near target – including floor
• Experiment – activate sample of vault concrete & analyze by γ spectroscopy • Reactor neutron spectrum not quite the same as cyclotron neutron spectrum • Your concrete may have different impurities
For both problems, suggest borated poly around target
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Questions?
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References [1] Mendez R. et. al. “Study of the neutron field around a PET cyclotron”. IRPA11 conference presentation,
May 2004. [2] Debeau J. et. al. “The Measurement of Neutron Energy Spectra in the High Neutron Flux Environments of Medical Accelerators Using Nested Neutron Spectrometer” (2013 CRPA conference poster). [3] Ar-41 source modeled as uniform equivalent sphere. Specific Gamma constant from Delacroix D. et al., Rad. Prot. Dosimetry v. 98, no. 1 pp. 9-18 (2002). [4] Thériault B. “Inhalation dose
coefficient for Ar-41”, private com. (2012). [5] β skin dose coefficient linearly extrapolated from data of - Fell TP, Rad. Prot. Dosimetry V. 36 No. 1, pp.
31-35 (1991). [6] Sunderland J. et al. “Considerations, measurements and logistics associated with low energy cyclotron
decommissioning (2011)”, AIP Conf. Proc. 1509, 16 (2012): http://dx.doi.org/10.1063/1.4773931. [7] Zagar T., Ravnik M., “Measurement of Neutron Activation In Concrete Samples”, Proc. Int. Conf.
“Nuclear Energy In Central Europe 2000”.
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Initial Results • 3m x 5m x 5m empty vault with point source
Ar40 Capture Spectrum in Vault air
0
0.000005
0.00001
0.000015
0.00002
0.000025
0.00003
1.E-09 1.E-07 1.E-05 1.E-03 1.E-01 1.E+01
Neutron Energy (MeV)A
r41 c
ap
ture
s (
per
so
urc
e n
eu
tro
n)
O16(n,p) Spectrum in Vault Air
0
1E-14
2E-14
3E-14
4E-14
5E-14
6E-14
7E-14
9.0E+00 1.0E+01 1.1E+01 1.2E+01 1.3E+01 1.4E+01 1.5E+01 1.6E+01
Neutron Energy (MeV)
O16
(n,p
) (pe
r so
urce
neu
tron
)