Activation of Air and Concrete in Medical Isotope Production Cyclotron Facilities CRPA 2016, Toronto Adam Dodd Senior Project Officer Accelerators and Class II Prescribed Equipment Division (613) 993-7930 or adam.dodd@canada.ca nuclearsafety.gc.ca
Outline Motivation Methods Air activation Concrete activation Conclusions 2
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! 3
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 4
Methods (2/2) Collaboration with UOIT co-op students Rob Shackelton (air activation) Devon Carr (concrete activation) Audrie Ismail (concrete activation) 5
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 6
Geometry and Materials Go from complex to extremely simple Reveals essential features and save computing time Concrete 6 m Iron Void Air Neutron source 3 m 2 m x y 12 m 7
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 8
Neutron Source Spectra (2/2) 9
Neutron Source Placement 3m x 5m x 5m vault 1.00% Effect of Source Placement on Ar-41 Production Eight source positions Starting at origin, moving diagonally into corner Considering statistical error, effect of source location is negligible Percent difference from mean Ar-41 production 0.50% 0.00% -0.50% -1.00% -1.50% 0 0.5 1 1.5 2 2.5 3 3.5 Distance from Origin (m) 10
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 Ar-40 Capture density 0 1 2 3 4 5 6 7 8 Cube edge length (m) 11
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/cm 3 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 12
Target Clamshell Polyethylene target shielding 5, 10, 50cm thickness No boron Concrete Air Source Critical thickness Thin shield thermalizes neutrons >10cm necessary to capture x 6 m Iron Void 2 m 12 m Polyethylene clamshell 3 m y γ shielding requirements 13
Partition Walls Wall placement below resulted in 15% decrease in Ar-41 production Concrete Objects in vault reduce air activation Justifies simplified geometry 7 m Iron Void Air 2 m Source 3 m x 20 m y 14
Ar-41 Activity Results are in captures/neutron - How many neutrons per μa of beam? IAEA TRS-468 x 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 15
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 gamma 3 5.1 17.3 Inhalation 4 0.1 0.3 Skin 5 0.9 2.2 Assumes saturation production Skin dose upper limit assumes no clothing 16
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 17
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? 18
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 τ ½ > 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 UCL: unconditional clearance level at which material can be thrown out as non-radioactive, * Made by fast neutrons via (n,p) reaction rather than (n,γ) 19
Literature Review (2/2) Reactor study [7] Ordinary concrete sample in TRIGA reactor in Slovenia 30 minute exposure @ neutron flux of 6.8 10 12 n/s/cm 2 Principal activities found Eu-152 and Co-60 at 6 Bq/g Conclusion concrete activation could be a problem with the new cyclotrons 20
Absorption depth Neutron capture density in concrete (F-18 production) Neutron capture density No Polyethylene 10 cm thick Polyethylene 10 cm thick Borated Polyethylene 1 3 5 15 25 35 45 55 65 80 100 Concrete depth (cm) 21
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 22
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 23
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) 24
2. Moving the Source Position on Y-Axis With 20-cm Thick Polyethylene Layer 75 250 425 600cm 25
Capture Density relative to when the source is in middle position 3.5 3.0 2.5 2.0 1.5 1.0 0.5 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 0 100 200 300 400 500 600 700 Distance between source and left wall (cm) 26
3. Lateral Distribution of Neutron Capture Density in Near Wall Capture density in the front wall 250 200 150 100 50 0-50 0 50 100 150 200 250 300 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%) 27
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 28
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 For both problems, suggest borated poly around target 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 29
Questions? 30
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. 31
Initial Results 3m x 5m x 5m empty vault with point source Ar40 Capture Spectrum in Vault air O16(n,p) Spectrum in Vault Air 0.00003 7E-14 0.000025 0.00002 0.000015 0.00001 0.000005 Ar41 captures (per source neutron) O16(n,p) (per source neutron) 6E-14 5E-14 4E-14 3E-14 2E-14 1.E-09 1.E-07 1.E-05 1.E-03 1.E-01 1.E+01 Neutron Energy (MeV) 0 1E-14 0 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) 32