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

2 Canadian Nuclear Safety Commission

Outline

• Motivation • Methods • Air activation • Concrete activation • Conclusions


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!


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


Methods (2/2)

Collaboration with UOIT co-op students

– Rob Shackelton (air activation) – Devon Carr (concrete activation) – Audrie Ismail (concrete activation)


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


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


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


Neutron Source Spectra (2/2)


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


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


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


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


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


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


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


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


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?


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,


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


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


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


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


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)


2. Moving the Source Position on Y-Axis With 20-cm Thick Polyethylene Layer

75 250 425 600cm


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)


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%)


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


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


Questions?


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”.

http://dx.doi.org/10.1063/1.4773931


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

)

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Activation of Air and Concrete in Medical Isotope ... · “The Measurement of Neutron Energy...

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