- Email: [email protected]
The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities
Applied Radiation and Isotopes 70 (2012) 620–624
Contents lists available at SciVerse ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities J. Ghassoun n, N. Senhou EPRA, Department of Physics, Faculty of Sciences Semlalia, PO Box: 2390, 40000 Marrakech, Morocco
a r t i c l e i n f o
a b s t r a c t
Article history: Received 15 July 2011 Received in revised form 21 December 2011 Accepted 22 December 2011 Available online 30 December 2011
In this study, the MCNP5 code was used to model radiotherapy room of a medical linear accelerator operating at 18 MV and to evaluate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions inside a liquid tissue-equivalent (TE) phantom. The obtained results were compared with measured data published in the literature. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall were also investigated. Our simulation results showed that paraffin wax containing boron carbide presents enough effectiveness to reduce both neutron and secondary gamma ray doses. & 2011 Elsevier Ltd. All rights reserved.
Keywords: Radiotherapy Tissue Dose Photoneutrons Shielding MCNP
1. Introduction Medical linear accelerators operating above 10 MV produce photoneutrons via interaction of high-energy photons with high atomic number (Z) materials. These photoneutrons which contaminate the therapeutic beam can also generate secondary gamma rays, via inelastic and capture reactions, which increase the undesirable dose delivered to the patient, the oncology staff and the general public. It is therefore advisable to have an accurate evaluation of neutron dose produced in the radiotherapy room, and delivered not only on the patient inside the treatment room but also in the surrounding areas. In a previous work (Ghassoun et al., 2011), the neutron and the secondary gamma ray dose equivalents were evaluated at various points inside the treatment room and along the maze without patient. In the present study, the Monte Carlo method has been used to model a radiotherapy room of a medical linear accelerator operating at 18 MV and to estimate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions in a tissue equivalent phantom representing a patient body. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall have been investigated in order to find the best shield to minimize the
n
Corresponding author. Fax: þ212 5 24 43 74 10. E-mail address: [email protected] (J. Ghassoun).
0969-8043/$ - see front matter & 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2011.12.041
undesirable neutron and secondary gamma ray doses delivered to the patient. The simulations were performed using the Monte Carlo N-Particle transport code (MCNP5) (X-5, 2005), in its coupled neutron–photon transport mode. The MCNP solid-state S(a,b) neutron scattering library (lwtr.01 t) was used in order to improve the accuracy of low energy neutron transport calculations. In our Monte Carlo simulation, 5 108 histories were run to obtain an estimate relative error of less than 1%.
2. Materials and methods 2.1. Monte Carlo simulation The application of the Monte Carlo method in medical physics covers almost all topics, including radiation protection, diagnostic, radiotherapy and nuclear medicine. In this study, the Monte Carlo method was used to model a radiotherapy room of a medical linear accelerator operating at 18 MV and to study the spatial-energy distribution of neutrons and secondary gamma rays in a 100 30 20 cm3 parallelepiped liquid tissue-equivalent (TE) phantom representing a patient body (International Commission on Radiation Units and Measurements (ICRU), 1989). Fig. 1 shows a cross view of the simulated radiotherapy room, the tissue equivalent phantom and the accelerator head. We assumed that all walls including the floor and the ceiling are made of ordinary concrete (National Council on Radiation
J. Ghassoun, N. Senhou / Applied Radiation and Isotopes 70 (2012) 620–624
Spherical tungsten shell
not be neglected because of the additional doses that patients can receive. To reduce the undesirable doses to the patient, the neutron and gamma ray energy spectra and dose equivalent distributions in the tissue phantom were also simulated when different neutron material shields in lining the radiotherapy room walls were used. Shielding effects were performed through the analysis of neutron and gamma ray depth dose curves and lateral dose profiles. The investigated neutron shielding materials (hydrocarbon class) includes: wood, paraffin wax containing borated carbide (3% B4C), lithiated polyethylene and PremadexTM, based on an organo-lithium salt (Waller et al., 2003).
Neutron source SSD =100
2.7m Z Y
621
Phantom 6m
3. Results and discussion
Fig. 1. A cross view of the simulated radiotherapy room, the tissue equivalent phantom and the accelerator head.
Protection and Measurements (NCRP), 1976). The distance from the floor to the ceiling was 2.7 m. The Monte Carlo N-Particle MCNP5 code was used to calculate the particle energy spectra, the fluence and dose equivalent of all components including fast neutrons, epithermal neutrons, thermal neutrons, and induced gamma rays at various points inside the tissue equivalent phantom. The thermal neutron region was defined to be below 0.5 eV, the epithermal neutron region is from 0.5 eV to 10 keV, and the fast neutron region is over 10 keV. Particle fluences were tallied in the phantom using the track length estimator (tally type F4), which calculates the mean fluence of particles in a cell volume. In this work, all calculations were done for a field size of 10 cm 10 cm at source-to-surface (SSD) of 100 cm. In this simulation, the accelerator head has been modeled as a 10 cm radius tungsten sphere around the source where neutrons are produced. The source is described as an isotropic point-like source energy spectrum given by Eq. (1) (Tosi et al., 1991): dN 0:8929En 0:1071lnðEmax =ðEn þ 7:34ÞÞ ¼ expðEn =TÞ þ R E 7:34 max dEn T2 lnðEmax =ðEn þ7:34ÞÞdEn 0 ð1Þ where Emax is the maximum energy of the photons (in MeV), En is the energy of neutrons (in MeV), and T is the nuclear temperature (in MeV) of the target material. In order to compare our simulated neutron dose equivalent with experimentally measured data, a value of Qn ¼ 1.5 1012 (n/ Gy X-ray) was used in this study (McGinley, 1998). Qn is the apparent neutron source strength in neutrons emitted from the accelerator head per Gy of X-rays absorbed at the isocenter. To estimate the neutron dose equivalents, the calculated neutron fluences were convoluted with updated neutron kerma coefficients (Chadwick et al., 1999) and the neutron quality factors based on the ICRU stopping power data for protons and alpha particles (Siebert and Schumacher, 1995). The neutron dose equivalent (H) was determined using the following equation: X H¼ jðEÞKðEÞQ ðEÞ ð2Þ E
where K(E) represents the neutron kerma factors and Q(E) represents the neutron quality factors and j(E) is the neutron fluence. For photons, the dose equivalent was calculated using mass energy absorption coefficients (Hubbell and Seltzer, 1997).
The spatial distribution of the epithermal neutron, thermal neutron, fast neutron, and gamma ray fluences along the central axis of the primary beam as a function of the depth in tissue phantom are plotted in Fig. 2. The fluence values are normalized per neutron emitted from the source. The fast and epithermal neutrons decrease with increasing the distance in the tissue phantom due to slowing down and moderation. This moderation gives rise to thermal neutrons, which reach a maximum fluence, at approximately 3 cm depth in the tissue phantom after what it decreases sharply. This figure also illustrates the secondary gamma rays fluence which is produced from (n,g) reaction with structure materials and within the tissue phantom component. Fig. 3 shows the dose equivalents components as a function of depth in tissue phantom along the main axis of the beam. Each dose component was calculated in small cylindrical tally volume cells along the beam centreline. The fast and epithermal neutrons dose equivalents fall off quickly due to rapid thermalization by light elements such hydrogen, while thermal neutron dose equivalent shows an increase with depth. The curve representing the gamma ray depth dose equivalents is similar in shape to thermal neutron dose equivalent since both are a direct result of the magnitude of the thermalization process of neutrons in the tissue phantom. However, the induced gamma ray dose equivalents decrease slightly at larger depths than the thermal neutrons dose equivalent due to the larger path of gamma rays compared to neutrons. This figure also indicates that the fast neutrons contribute a substantially higher dose than the other dose components and its maximum occurs at the tissue phantom surface. Fig. 4 shows the equivalent dose (Sv), from neutrons per unit absorbed dose (Gy) from primary X-rays, as a function of energy 8,E-06 Thermal neutrons Epithermal neutrons 7,E-06 Fast neutrons Gamma rays 6,E-06
3.0E-05 2.5E-05 2.0E-05
5,E-06 4,E-06
1.5E-05
3,E-06
1.0E-05
2,E-06 5.0E-06
1,E-06
0.0E+00
2.2. Shielding effets and dose equivalent reduction The dose equivalent due to neutrons produced by medical linear accelerators operating above 10 MV is relevant and should
0
2
4
6
8 10 12 Depth (cm)
14
16
18
0,E+00 20
Gamma rays fluence (n/cm2) per source neutron
1m
Neutron fluence (n/cm2) per source neutron
Concrete
Fig. 2. The simulated neutron and the gamma ray fluence components as a function of depth in tissue phantom.
J. Ghassoun, N. Senhou / Applied Radiation and Isotopes 70 (2012) 620–624
1,E+03
1,E-02 Thermal neutrons Epithermal neutrons Fast neutrons Gamma rays
1,E+02 Quality Factor
1,E-04
1,E+01
1,E+00
1,E+00
1,E-01
1,E-01
1,E-02
1,E-02
1,E-03
1,E-03 -0 4 1, E03 1, E02 1, E0 1, 1 E+ 0 1, 0 E+ 0 1, 1 E+ 02
-0 5
Energy (MeV)
1,E-07 0
2
4
6
8
10 12 Depth (cm)
14
16
18
20
Fig. 3. The simulated dose equivalent components, in logarithmic scale, as a function of depth in the tissue phantom.
1,E-03 1,E-04
1, E
1, E
-0 6
-0 7
1, E
-0 9
1, E
1, E
-0 8
1,E-05
1,E-06
0.25 cm depth 5 cm depth 10 cm depth
Fig. 5. The neutron kerma and quality factor for the used liquid tissue equivalent phantom as a function of energy.
Table 1 Comparison between the calculated neutron dose equivalents and the measured ones reported by d’Errico at a depth of 1 cm in the liquid tissue equivalent phantom. Distance from the central Neutron dose equivalents per unit photon dose beam axis (cm) (mSv/Gy) Our work
d’Errico et al. (1998)
RD (%)
4.55 70.0114 2.01 70.0165 1.78 70.0173
4.50 2.05 1.75
1.11 1.99 1.71
1,E-05 0 10 20
1,E-06 1,E-07
2,E-06 2,E-06 1,E-06 5,E-07
+0 0 E+ 01 1,
01
1, E
E-
02
1,
E-
03
1,
04
E1,
1,
E-
05 E-
06
1,
07
E-
1,
1,
E-
08
0,E+00
E-
calculated at different depths in the tissue equivalent phantom. The contribution of high-energy neutrons (between 100 keV and 1 MeV) to the total neutron dose equivalent is dominant in all spectra which confirms the findings stated previously. The variation of the neutron kerma K(E), the quality factor Q(E), and the product K(E) Q(E) with energy for the used tissue equivalent phantom is plotted in Fig. 5. This figure clearly indicates that fast neutrons present higher kerma conversion and quality factor values confirming their larger contribution to the total neutron dose equivalents. To check the validity of our Monte Carlo simulation, we have compared the results of the current study with experimentally measured data published in the literature. Table 1 shows the comparison between our simulated and measured neutron dose equivalent at a depth of 1 cm in the tissue phantom for different distances from the central beam axis. The experimentally measured values were performed by using a superheated emulsion (SE) chamber detector (d’Errico et al., 1998). There is a good agreement between the MCNP calculations and the published measurements and the relative difference (RD) between the two data sets are less than 2%. Relative errors, i.e, the standard deviation of the mean divided by the mean, calculated by MCNP5 code associated to simulation results are less than 1%. Fig. 6 shows the behavior of the neutron energy spectra calculated at the surface of the tissue phantom (0.25 cm depth),
3,E-06
1,
Fig. 4. The neutron dose equivalent as a function of energy calculated at different depths in the tissue equivalent phantom.
Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
09
Energy (MeV)
3,E-06
E-
1,E-09 1,E-08 1,E-07 1,E-06 1,E-05 1,E-04 1,E-03 1,E-02 1,E-01 1,E+00 1,E+01
1,
1,E-08
Neutron fluence (n cm-2) per source neutron
Neutron dose equivalent (Sv/Gy x-ray dose)
1,E+02
1,E+01
1, E
Dose equivalents (Sv/Gy)
1,E-03
1,E+03 Quality factor Q Product KxQ Kerma K
Neutron kerma (pGy cm2)
622
Energy (MeV) Fig. 6. The neutron energy spectra calculated at a depth of 0.25 cm, in the tissue phantom, in the presence of neutron material shields.
in the presence of neutron material shields of thickness 2.5 cm lining the radiotherapy room walls. This result reflects the differences in neutron and gamma ray scattering properties and albedo between the different materials. The neutron fluence is more reduced when the room walls were lined with hydrocarbon neutron shields containing admixtures of neutron capture materials such as boron or lithium. The addition of lithium or boron to hydrogen materials enhances the thermal neutron capture because of their high thermal-neutron capture cross sections (Ghassoun and Mostacci, 2011). The natural boron carbide loaded paraffin wax provides a minimum amount of fast and thermal
Neeutron dose equivalent (mSv/Gy)
5.15E+00
Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
4.65E+00 4.15E+00
Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
4,7E+00 4,2E+00 3,7E+00 3,2E+00 2,7E+00 2,2E+00
-10 0 10 20 -20 Distance from the central beam axis (cm)
30
Fig. 9. Neutron dose equivalent profiles at 0.25 cm depth in the tissue phantom for different material lining the radiotherapy room walls.
4.0E-02 3.5E-02 3.0E-02 2.5E-02 Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
2.0E-02 1.5E-02 -30
3.65E+00
20 -20 -10 0 10 Distance from the central beam axis (cm)
30
Fig. 10. Prompt gamma-ray dose equivalent profiles at 0.25 cm depth in the tissue phantom for different materials lining the room walls.
3.15E+00 2.65E+00
Table 2 Total dose equivalents (H) at a depth of 0.25 cm in the tissue phantom for different types of neutron shielding materials lining the ordinary concrete walls.
2.15E+00 0
0.5
1
1.5
2 2.5 Depth (cm)
3
3.5
4 Wall composition
Total dose (neutron plus gamma ray) equivalents (mSv/Gy)
Factor of reduction (%)
Concrete Concrete þ2.5 cm wood Concrete þ2.5 cm paraffin
5.14 5.03 4.93
2.06 4.12
4.97
3.37
4.95
3.75
Fig. 7. The simulated neutron depth dose equivalents in the tissue phantom along the central beam axis.
Gamma-rays dose equivalent (mSv/Gy)
623
5,2E+00
1,7E+00 -30
Gamma-ray dose equivalents (mSv/Gy)
neutrons and consequently provides the lowest damage of surface tissues in comparison to other material shields. Fig. 7 shows the calculated total neutron dose equivalent distributions along the central beam axis as a function of depth in the tissue phantom. The obtained results show that the use of hydrocarbon neutron shields of thickness 2.5 cm, containing admixtures of neutron capture materials lining radiotherapy room walls, can significantly reduce the undesirable neutron dose equivalents to the patient. The neutron shielding effects on the neutron dose equivalents are more pronounced at the surface of the tissue phantom (for depths o3 cm). The natural boron carbide loaded paraffin wax shows enough effectiveness to reduce the neutron dose delivered to the tissue equivalent phantom simulating a patient body. Fig. 8 shows the calculated gamma ray dose equivalent distributions along the central beam axis as a function of depth in the tissue phantom, in the presence of neutron material shields of thickness 2.5 cm lining the radiotherapy room walls. The hydrocarbon neutron shields containing lithium like Premadex or lithium doped polyethylene, are more effective in reducing secondary gamma-ray dose equivalents. The lithium does not produce capture gamma rays and the absorption is predominantly due to (n,a) reaction. But lithium is less effective than boron for absorbing thermal neutrons. Fig. 9 shows the total neutron dose equivalent profiles calculated at a depth of 0.25 cm in the tissue phantom in a direction perpendicular to the central axis. It can be seen that there is a
Neutron dose equivalent (mSv/Gy)
J. Ghassoun, N. Senhou / Applied Radiation and Isotopes 70 (2012) 620–624
5.5E-02
Wax containing boron Concrete þ2.5 cm lithiated Polyethylene Concrete þ2.5 cm Premadex
5.0E-02 4.5E-02 4.0E-02 3.5E-02 3.0E-02 2.5E-02
Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
2.0E-02 1.5E-02 1.0E-02 0
2
4
6
8 10 Depth (cm)
12
14
16
18
Fig. 8. The gamma-ray depth dose equivalents calculated in the tissue phantom along the central beam axis.
significant reduction in neutron dose equivalent when the room walls were lined with paraffin wax containing boron carbide. Fig. 10 shows the secondary gamma rays dose equivalent profiles calculated by Monte Carlo, at a depth of 0.25 cm in the tissue phantom when different materials in lining the radiotherapy room walls were used. The greatest reduction is found when lithiated polyethylene is used to line the ordinary concrete walls. Table 2 summarizes the simulated total (neutron plus secondary gamma ray) dose equivalents obtained at a depth of 0.25 cm in the tissue phantom, for different types of neutron shielding material lining the ordinary concrete walls. It is clear from this table that paraffin wax containing boron carbide exhibits the best characteristic for reducing both neutron and prompt gamma ray
624
J. Ghassoun, N. Senhou / Applied Radiation and Isotopes 70 (2012) 620–624
dose equivalents. It reduces the total dose equivalents at surface tissue phantom by 4.12%.
described here could be easily extended to other medical accelerators as well.
4. Conclusion
References
In this paper, the Monte Carlo N-Particle MCNP5 code was used to model the radiotherapy room for a medical linear accelerator operating at 18 MV and to evaluate the neutron and secondary gamma ray dose equivalent distributions inside a tissue-equivalent liquid phantom simulating a patient’s body. A comparison of Monte Carlo results with experimental data available in the literature is also presented. The obtained results showed that the neutrons and seondary gamma rays that contaminate the therapeutic beam give a non-negligible contribution particularly at the tissue phantom surface. Also, it was observed the fast neutrons contribute a substantially higher dose than the other dose components and its maximum occurs at the tissue phantom surface. In addition, the effect of neutron material shields on the radiotherapy room walls has also been investigated. According to MCNP results, it was determined that the total (neutron plus gamma-ray) dose equivalent at the entrance of tissue equivalent phantom is reduced by 4.12% using customised neutron shield as compared with a facility not using this shielding. According to the current study, we conclude that paraffin wax containing boron carbide appears to have a more optimal design for shielding both neutron and prompt gamma rays. The data presented in this study could provide more and accurate information on the dose equivalent distributions in a patient and also data assisting in the optimization of radiation shielding of a high energy medical accelerator room. The simulated linear accelerator was a Saturne 20 (CGR) but the method
Chadwick, M.B., Barschall, H.H., Caswell, R.S., et al., 1999. A consistent set of neutron kerma coefficients from thermal to 150MeV for biologically important materials. Med. Phys. 26, 974. d’Errico, F., Nath, R., Tana, L., Corzio., G., Alberts, W.G., 1998. In-phantom dosimetry and spectrometry of photoneutrons from an 18 MV linear accelerator. Med. Phys. 25, 1717–1724. Ghassoun, J., Senhou, N., Jehouani, A., 2011. Neutron and photon doses in high energy radiotherapy facilities and evaluation of shielding performance by Monte Carlo method. Ann. Nucl. Energy 38, 2163–2167. Ghassoun, J., Mostacci, D., 2011. A compact neutron beam generator system designed for prompt gamma nuclear activation analysis. Appl. Radiat. Isot. 69 (8), 1138–1142. Hubbell, J.H., Seltzer, S.M., 1997. Tables of X-ray Mass Attenuation Coefficients and Mass Energy–Absorption Coefficients. National Institute of Standards and Technology, Gaithersburg, MD. /http:/physics.nist.gov/xaamdiS. International Commission on Radiation Units and Measurements (ICRU)., 1989. Tissue Substitutes in Radiation Dosimetry and Measurement. ICRU-44. McGinley, P.H., 1998. Shielding Techniques for Radiation Oncology Facilities. Medical Physics Publishing, Madison. WI. National Council on Radiation Protection and Measurements (NCRP)., 1976. Structural shielding design and evaluation for medical use of X-rays and gamma rays of energies up to 10 MeV. NCRP Report 49. National Council on Radiation Protection and Measurements. Bethesda. MD. USA. Siebert, B.R.L., Schumacher, H., 1995. Quality factors, ambient and personal dose equivalent for neutrons based on the new ICRU stopping power data for protons and alpha particles. Radiat. Prot. Dosim. 58 (3), 177–183. Tosi, G., Torresin, A., Agosteo, S., Foglio Para, A., Sangiust, V., Zeni, L., Silari, M., 1991. Neutron measurements around medical electron accelerators by active and passive detection techniques. Med. Phys. 18 (1), 54–60. Waller, E.J., Jamieson, T.J., Cole, D., Cousins, T., Jammal, R.B., 2003. Effectiveness of custom neutron shielding in the maze of radiotherapy accelerators. Radiat. Prot. Dosim. 107 (4), 233–238. X-5 Monte Carlo Team, 2005. MCNP – A General Monte Carlo N-Particle Transport Code. Version 5. Los Alamos National Laboratory report LAUR- 03-1987 (April 2003; Revised October 3. 2005).
Contents lists available at SciVerse ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities J. Ghassoun n, N. Senhou EPRA, Department of Physics, Faculty of Sciences Semlalia, PO Box: 2390, 40000 Marrakech, Morocco
a r t i c l e i n f o
a b s t r a c t
Article history: Received 15 July 2011 Received in revised form 21 December 2011 Accepted 22 December 2011 Available online 30 December 2011
In this study, the MCNP5 code was used to model radiotherapy room of a medical linear accelerator operating at 18 MV and to evaluate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions inside a liquid tissue-equivalent (TE) phantom. The obtained results were compared with measured data published in the literature. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall were also investigated. Our simulation results showed that paraffin wax containing boron carbide presents enough effectiveness to reduce both neutron and secondary gamma ray doses. & 2011 Elsevier Ltd. All rights reserved.
Keywords: Radiotherapy Tissue Dose Photoneutrons Shielding MCNP
1. Introduction Medical linear accelerators operating above 10 MV produce photoneutrons via interaction of high-energy photons with high atomic number (Z) materials. These photoneutrons which contaminate the therapeutic beam can also generate secondary gamma rays, via inelastic and capture reactions, which increase the undesirable dose delivered to the patient, the oncology staff and the general public. It is therefore advisable to have an accurate evaluation of neutron dose produced in the radiotherapy room, and delivered not only on the patient inside the treatment room but also in the surrounding areas. In a previous work (Ghassoun et al., 2011), the neutron and the secondary gamma ray dose equivalents were evaluated at various points inside the treatment room and along the maze without patient. In the present study, the Monte Carlo method has been used to model a radiotherapy room of a medical linear accelerator operating at 18 MV and to estimate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions in a tissue equivalent phantom representing a patient body. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall have been investigated in order to find the best shield to minimize the
n
Corresponding author. Fax: þ212 5 24 43 74 10. E-mail address: [email protected] (J. Ghassoun).
0969-8043/$ - see front matter & 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2011.12.041
undesirable neutron and secondary gamma ray doses delivered to the patient. The simulations were performed using the Monte Carlo N-Particle transport code (MCNP5) (X-5, 2005), in its coupled neutron–photon transport mode. The MCNP solid-state S(a,b) neutron scattering library (lwtr.01 t) was used in order to improve the accuracy of low energy neutron transport calculations. In our Monte Carlo simulation, 5 108 histories were run to obtain an estimate relative error of less than 1%.
2. Materials and methods 2.1. Monte Carlo simulation The application of the Monte Carlo method in medical physics covers almost all topics, including radiation protection, diagnostic, radiotherapy and nuclear medicine. In this study, the Monte Carlo method was used to model a radiotherapy room of a medical linear accelerator operating at 18 MV and to study the spatial-energy distribution of neutrons and secondary gamma rays in a 100 30 20 cm3 parallelepiped liquid tissue-equivalent (TE) phantom representing a patient body (International Commission on Radiation Units and Measurements (ICRU), 1989). Fig. 1 shows a cross view of the simulated radiotherapy room, the tissue equivalent phantom and the accelerator head. We assumed that all walls including the floor and the ceiling are made of ordinary concrete (National Council on Radiation
J. Ghassoun, N. Senhou / Applied Radiation and Isotopes 70 (2012) 620–624
Spherical tungsten shell
not be neglected because of the additional doses that patients can receive. To reduce the undesirable doses to the patient, the neutron and gamma ray energy spectra and dose equivalent distributions in the tissue phantom were also simulated when different neutron material shields in lining the radiotherapy room walls were used. Shielding effects were performed through the analysis of neutron and gamma ray depth dose curves and lateral dose profiles. The investigated neutron shielding materials (hydrocarbon class) includes: wood, paraffin wax containing borated carbide (3% B4C), lithiated polyethylene and PremadexTM, based on an organo-lithium salt (Waller et al., 2003).
Neutron source SSD =100
2.7m Z Y
621
Phantom 6m
3. Results and discussion
Fig. 1. A cross view of the simulated radiotherapy room, the tissue equivalent phantom and the accelerator head.
Protection and Measurements (NCRP), 1976). The distance from the floor to the ceiling was 2.7 m. The Monte Carlo N-Particle MCNP5 code was used to calculate the particle energy spectra, the fluence and dose equivalent of all components including fast neutrons, epithermal neutrons, thermal neutrons, and induced gamma rays at various points inside the tissue equivalent phantom. The thermal neutron region was defined to be below 0.5 eV, the epithermal neutron region is from 0.5 eV to 10 keV, and the fast neutron region is over 10 keV. Particle fluences were tallied in the phantom using the track length estimator (tally type F4), which calculates the mean fluence of particles in a cell volume. In this work, all calculations were done for a field size of 10 cm 10 cm at source-to-surface (SSD) of 100 cm. In this simulation, the accelerator head has been modeled as a 10 cm radius tungsten sphere around the source where neutrons are produced. The source is described as an isotropic point-like source energy spectrum given by Eq. (1) (Tosi et al., 1991): dN 0:8929En 0:1071lnðEmax =ðEn þ 7:34ÞÞ ¼ expðEn =TÞ þ R E 7:34 max dEn T2 lnðEmax =ðEn þ7:34ÞÞdEn 0 ð1Þ where Emax is the maximum energy of the photons (in MeV), En is the energy of neutrons (in MeV), and T is the nuclear temperature (in MeV) of the target material. In order to compare our simulated neutron dose equivalent with experimentally measured data, a value of Qn ¼ 1.5 1012 (n/ Gy X-ray) was used in this study (McGinley, 1998). Qn is the apparent neutron source strength in neutrons emitted from the accelerator head per Gy of X-rays absorbed at the isocenter. To estimate the neutron dose equivalents, the calculated neutron fluences were convoluted with updated neutron kerma coefficients (Chadwick et al., 1999) and the neutron quality factors based on the ICRU stopping power data for protons and alpha particles (Siebert and Schumacher, 1995). The neutron dose equivalent (H) was determined using the following equation: X H¼ jðEÞKðEÞQ ðEÞ ð2Þ E
where K(E) represents the neutron kerma factors and Q(E) represents the neutron quality factors and j(E) is the neutron fluence. For photons, the dose equivalent was calculated using mass energy absorption coefficients (Hubbell and Seltzer, 1997).
The spatial distribution of the epithermal neutron, thermal neutron, fast neutron, and gamma ray fluences along the central axis of the primary beam as a function of the depth in tissue phantom are plotted in Fig. 2. The fluence values are normalized per neutron emitted from the source. The fast and epithermal neutrons decrease with increasing the distance in the tissue phantom due to slowing down and moderation. This moderation gives rise to thermal neutrons, which reach a maximum fluence, at approximately 3 cm depth in the tissue phantom after what it decreases sharply. This figure also illustrates the secondary gamma rays fluence which is produced from (n,g) reaction with structure materials and within the tissue phantom component. Fig. 3 shows the dose equivalents components as a function of depth in tissue phantom along the main axis of the beam. Each dose component was calculated in small cylindrical tally volume cells along the beam centreline. The fast and epithermal neutrons dose equivalents fall off quickly due to rapid thermalization by light elements such hydrogen, while thermal neutron dose equivalent shows an increase with depth. The curve representing the gamma ray depth dose equivalents is similar in shape to thermal neutron dose equivalent since both are a direct result of the magnitude of the thermalization process of neutrons in the tissue phantom. However, the induced gamma ray dose equivalents decrease slightly at larger depths than the thermal neutrons dose equivalent due to the larger path of gamma rays compared to neutrons. This figure also indicates that the fast neutrons contribute a substantially higher dose than the other dose components and its maximum occurs at the tissue phantom surface. Fig. 4 shows the equivalent dose (Sv), from neutrons per unit absorbed dose (Gy) from primary X-rays, as a function of energy 8,E-06 Thermal neutrons Epithermal neutrons 7,E-06 Fast neutrons Gamma rays 6,E-06
3.0E-05 2.5E-05 2.0E-05
5,E-06 4,E-06
1.5E-05
3,E-06
1.0E-05
2,E-06 5.0E-06
1,E-06
0.0E+00
2.2. Shielding effets and dose equivalent reduction The dose equivalent due to neutrons produced by medical linear accelerators operating above 10 MV is relevant and should
0
2
4
6
8 10 12 Depth (cm)
14
16
18
0,E+00 20
Gamma rays fluence (n/cm2) per source neutron
1m
Neutron fluence (n/cm2) per source neutron
Concrete
Fig. 2. The simulated neutron and the gamma ray fluence components as a function of depth in tissue phantom.
J. Ghassoun, N. Senhou / Applied Radiation and Isotopes 70 (2012) 620–624
1,E+03
1,E-02 Thermal neutrons Epithermal neutrons Fast neutrons Gamma rays
1,E+02 Quality Factor
1,E-04
1,E+01
1,E+00
1,E+00
1,E-01
1,E-01
1,E-02
1,E-02
1,E-03
1,E-03 -0 4 1, E03 1, E02 1, E0 1, 1 E+ 0 1, 0 E+ 0 1, 1 E+ 02
-0 5
Energy (MeV)
1,E-07 0
2
4
6
8
10 12 Depth (cm)
14
16
18
20
Fig. 3. The simulated dose equivalent components, in logarithmic scale, as a function of depth in the tissue phantom.
1,E-03 1,E-04
1, E
1, E
-0 6
-0 7
1, E
-0 9
1, E
1, E
-0 8
1,E-05
1,E-06
0.25 cm depth 5 cm depth 10 cm depth
Fig. 5. The neutron kerma and quality factor for the used liquid tissue equivalent phantom as a function of energy.
Table 1 Comparison between the calculated neutron dose equivalents and the measured ones reported by d’Errico at a depth of 1 cm in the liquid tissue equivalent phantom. Distance from the central Neutron dose equivalents per unit photon dose beam axis (cm) (mSv/Gy) Our work
d’Errico et al. (1998)
RD (%)
4.55 70.0114 2.01 70.0165 1.78 70.0173
4.50 2.05 1.75
1.11 1.99 1.71
1,E-05 0 10 20
1,E-06 1,E-07
2,E-06 2,E-06 1,E-06 5,E-07
+0 0 E+ 01 1,
01
1, E
E-
02
1,
E-
03
1,
04
E1,
1,
E-
05 E-
06
1,
07
E-
1,
1,
E-
08
0,E+00
E-
calculated at different depths in the tissue equivalent phantom. The contribution of high-energy neutrons (between 100 keV and 1 MeV) to the total neutron dose equivalent is dominant in all spectra which confirms the findings stated previously. The variation of the neutron kerma K(E), the quality factor Q(E), and the product K(E) Q(E) with energy for the used tissue equivalent phantom is plotted in Fig. 5. This figure clearly indicates that fast neutrons present higher kerma conversion and quality factor values confirming their larger contribution to the total neutron dose equivalents. To check the validity of our Monte Carlo simulation, we have compared the results of the current study with experimentally measured data published in the literature. Table 1 shows the comparison between our simulated and measured neutron dose equivalent at a depth of 1 cm in the tissue phantom for different distances from the central beam axis. The experimentally measured values were performed by using a superheated emulsion (SE) chamber detector (d’Errico et al., 1998). There is a good agreement between the MCNP calculations and the published measurements and the relative difference (RD) between the two data sets are less than 2%. Relative errors, i.e, the standard deviation of the mean divided by the mean, calculated by MCNP5 code associated to simulation results are less than 1%. Fig. 6 shows the behavior of the neutron energy spectra calculated at the surface of the tissue phantom (0.25 cm depth),
3,E-06
1,
Fig. 4. The neutron dose equivalent as a function of energy calculated at different depths in the tissue equivalent phantom.
Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
09
Energy (MeV)
3,E-06
E-
1,E-09 1,E-08 1,E-07 1,E-06 1,E-05 1,E-04 1,E-03 1,E-02 1,E-01 1,E+00 1,E+01
1,
1,E-08
Neutron fluence (n cm-2) per source neutron
Neutron dose equivalent (Sv/Gy x-ray dose)
1,E+02
1,E+01
1, E
Dose equivalents (Sv/Gy)
1,E-03
1,E+03 Quality factor Q Product KxQ Kerma K
Neutron kerma (pGy cm2)
622
Energy (MeV) Fig. 6. The neutron energy spectra calculated at a depth of 0.25 cm, in the tissue phantom, in the presence of neutron material shields.
in the presence of neutron material shields of thickness 2.5 cm lining the radiotherapy room walls. This result reflects the differences in neutron and gamma ray scattering properties and albedo between the different materials. The neutron fluence is more reduced when the room walls were lined with hydrocarbon neutron shields containing admixtures of neutron capture materials such as boron or lithium. The addition of lithium or boron to hydrogen materials enhances the thermal neutron capture because of their high thermal-neutron capture cross sections (Ghassoun and Mostacci, 2011). The natural boron carbide loaded paraffin wax provides a minimum amount of fast and thermal
Neeutron dose equivalent (mSv/Gy)
5.15E+00
Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
4.65E+00 4.15E+00
Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
4,7E+00 4,2E+00 3,7E+00 3,2E+00 2,7E+00 2,2E+00
-10 0 10 20 -20 Distance from the central beam axis (cm)
30
Fig. 9. Neutron dose equivalent profiles at 0.25 cm depth in the tissue phantom for different material lining the radiotherapy room walls.
4.0E-02 3.5E-02 3.0E-02 2.5E-02 Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
2.0E-02 1.5E-02 -30
3.65E+00
20 -20 -10 0 10 Distance from the central beam axis (cm)
30
Fig. 10. Prompt gamma-ray dose equivalent profiles at 0.25 cm depth in the tissue phantom for different materials lining the room walls.
3.15E+00 2.65E+00
Table 2 Total dose equivalents (H) at a depth of 0.25 cm in the tissue phantom for different types of neutron shielding materials lining the ordinary concrete walls.
2.15E+00 0
0.5
1
1.5
2 2.5 Depth (cm)
3
3.5
4 Wall composition
Total dose (neutron plus gamma ray) equivalents (mSv/Gy)
Factor of reduction (%)
Concrete Concrete þ2.5 cm wood Concrete þ2.5 cm paraffin
5.14 5.03 4.93
2.06 4.12
4.97
3.37
4.95
3.75
Fig. 7. The simulated neutron depth dose equivalents in the tissue phantom along the central beam axis.
Gamma-rays dose equivalent (mSv/Gy)
623
5,2E+00
1,7E+00 -30
Gamma-ray dose equivalents (mSv/Gy)
neutrons and consequently provides the lowest damage of surface tissues in comparison to other material shields. Fig. 7 shows the calculated total neutron dose equivalent distributions along the central beam axis as a function of depth in the tissue phantom. The obtained results show that the use of hydrocarbon neutron shields of thickness 2.5 cm, containing admixtures of neutron capture materials lining radiotherapy room walls, can significantly reduce the undesirable neutron dose equivalents to the patient. The neutron shielding effects on the neutron dose equivalents are more pronounced at the surface of the tissue phantom (for depths o3 cm). The natural boron carbide loaded paraffin wax shows enough effectiveness to reduce the neutron dose delivered to the tissue equivalent phantom simulating a patient body. Fig. 8 shows the calculated gamma ray dose equivalent distributions along the central beam axis as a function of depth in the tissue phantom, in the presence of neutron material shields of thickness 2.5 cm lining the radiotherapy room walls. The hydrocarbon neutron shields containing lithium like Premadex or lithium doped polyethylene, are more effective in reducing secondary gamma-ray dose equivalents. The lithium does not produce capture gamma rays and the absorption is predominantly due to (n,a) reaction. But lithium is less effective than boron for absorbing thermal neutrons. Fig. 9 shows the total neutron dose equivalent profiles calculated at a depth of 0.25 cm in the tissue phantom in a direction perpendicular to the central axis. It can be seen that there is a
Neutron dose equivalent (mSv/Gy)
J. Ghassoun, N. Senhou / Applied Radiation and Isotopes 70 (2012) 620–624
5.5E-02
Wax containing boron Concrete þ2.5 cm lithiated Polyethylene Concrete þ2.5 cm Premadex
5.0E-02 4.5E-02 4.0E-02 3.5E-02 3.0E-02 2.5E-02
Concrete Premadex Lithiated polyethylene Wood Paraffin wax containing boron carbide
2.0E-02 1.5E-02 1.0E-02 0
2
4
6
8 10 Depth (cm)
12
14
16
18
Fig. 8. The gamma-ray depth dose equivalents calculated in the tissue phantom along the central beam axis.
significant reduction in neutron dose equivalent when the room walls were lined with paraffin wax containing boron carbide. Fig. 10 shows the secondary gamma rays dose equivalent profiles calculated by Monte Carlo, at a depth of 0.25 cm in the tissue phantom when different materials in lining the radiotherapy room walls were used. The greatest reduction is found when lithiated polyethylene is used to line the ordinary concrete walls. Table 2 summarizes the simulated total (neutron plus secondary gamma ray) dose equivalents obtained at a depth of 0.25 cm in the tissue phantom, for different types of neutron shielding material lining the ordinary concrete walls. It is clear from this table that paraffin wax containing boron carbide exhibits the best characteristic for reducing both neutron and prompt gamma ray
624
J. Ghassoun, N. Senhou / Applied Radiation and Isotopes 70 (2012) 620–624
dose equivalents. It reduces the total dose equivalents at surface tissue phantom by 4.12%.
described here could be easily extended to other medical accelerators as well.
4. Conclusion
References
In this paper, the Monte Carlo N-Particle MCNP5 code was used to model the radiotherapy room for a medical linear accelerator operating at 18 MV and to evaluate the neutron and secondary gamma ray dose equivalent distributions inside a tissue-equivalent liquid phantom simulating a patient’s body. A comparison of Monte Carlo results with experimental data available in the literature is also presented. The obtained results showed that the neutrons and seondary gamma rays that contaminate the therapeutic beam give a non-negligible contribution particularly at the tissue phantom surface. Also, it was observed the fast neutrons contribute a substantially higher dose than the other dose components and its maximum occurs at the tissue phantom surface. In addition, the effect of neutron material shields on the radiotherapy room walls has also been investigated. According to MCNP results, it was determined that the total (neutron plus gamma-ray) dose equivalent at the entrance of tissue equivalent phantom is reduced by 4.12% using customised neutron shield as compared with a facility not using this shielding. According to the current study, we conclude that paraffin wax containing boron carbide appears to have a more optimal design for shielding both neutron and prompt gamma rays. The data presented in this study could provide more and accurate information on the dose equivalent distributions in a patient and also data assisting in the optimization of radiation shielding of a high energy medical accelerator room. The simulated linear accelerator was a Saturne 20 (CGR) but the method
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