**2. Semiconductor spectrometry and its application for radiation characterization**

Semiconductor high-purity germanium (HPGe) detectors are most suitable for the investigation of gamma radiation spectra of unknown origin since the excellent energy resolution enables the exact identification of any radionuclide, which contributes to the radiation field in measured localization. Their use is, however, limited due to the need of liquid nitrogen cooling; therefore, scintillation or roomtemperature semiconductor detectors, both with limited resolution, are used instead.

The spectrometric system used in this study, as shown in **Figure 4**, consists of:


The carbon-composite entrance window enables the registration of low-energy photons (above 7 keV). The end-point energy of measured spectra has been set

*Use of Gamma Radiation Techniques in Peaceful Applications*

higher nominal accelerating potentials [35].

depends on beam nominal potential, as presented in **Figure 3**.

radiation types but occurring with different efficiency.

which (n,n'γ) and (n,γ) are the most commonly observed, for fast and thermalized neutrons, respectively. Every mechanism mentioned above could activate radionuclides. Nevertheless, the majority of induced radioactivity is found in construction materials of the accelerator head, mostly in heavy elements of collimation and beam shaping system. The contribution of particular elements of linac head in overall induced radioactivity is studied mostly with Monte Carlo simulations, as in: [29, 30]. However, gamma radiation spectrometry is a good tool for identification of particular radionuclides and their contribution in this phenomenon, for example, see: [31]. The apparent linac radioactivity depends on the localization of measuring point; therefore, the radiation hazard due to this phenomenon is different for patients and for the staff, with the dominant contribution of tungsten collimator or head casing, respectively [32]. Induced radioactivity has been also observed and investigated in tissues [33–36], air [37], treatment couch [38], and treatment accessories stored inside the linac room [39]. Moreover, the dependence of induced activity on the therapeutic dose rate could be observed in some cases, i.e., when half-life of radioisotope is comparable with the time of beam emission, and is more pronounced for

Among the mechanisms of radionuclide activation outside the field of irradiation, neutron capture contributes the most. Linacs used nowadays are not routinely equipped with shielding constructions dedicated for neutrons; therefore, neutron fluence all over the treatment room is reported [32, 40–44] in the amount sufficient for inducing radioactivity at measurable level. Therefore, medical linear accelerators are often characterized in terms of neutron source strength Q [14, 44], which

The spectrum of neutron flux undergoes changes via scattering mechanisms. Leaving the linac head, the mean energy of neutrons is of the order of 1 MeV, on treatment couch, an additional peak at thermal energies is already observed and neutrons impinging the door have an average energy of ~0.2 MeV [45, 46]. Neutron radiation weighting factor for effective dose calculation strongly depends on energy, having maximal values around 1 MeV [47]. The cross section of (n,γ) nuclear reaction follows the 1/E dependence with some resonance peaks at intermediate energies [28]. Therefore, high-energy neutrons contribute mostly to the dose, whereas slow neutrons to the phenomenon of induced radioactivity. It is of high importance to be aware of the physical mechanisms of radiation absorption and removal from the beam. These are in principle different for various radiation types. Nevertheless, similar mechanisms might be observed for various

The readily used in diagnostic radiology heavy metal shielding is no longer valid in high-energy radiotherapy rooms due to the generation of secondary

*The comparison of neutron source strength values reported in [14]* (0)*, [44]* (□)*, and obtained by us* (Δ)*.*

**162**

**Figure 3.**

#### **Figure 4.**

*ReGe gamma spectrometer used in present study: (a) multichannel analyzer, (b) front view of high-purity germanium detector with carbon-composite entrance window, and (c) measurement configuration 50 cm from the entrance door to the linac radiotherapy room.*

by adjusting the gain value of MCA. This enables for spectra registration up to the energy of 3–10 MeV.

Energy calibration of spectrometric system has been performed for standard MCA gain (5.0) with the use of radioisotope sealed sources (1 cm Ø) having activity of the order of 10 kBq. Subsequently, it was checked that the calibration scales linearly inversely with the spectrometric gain.

Spectrometric efficiency has been modeled in In Situ Object Counting Software (ISOCS™, Canberra Inc.), applying full factory characterization of a given detector performed with the use of NIST-traceable sources and MCNP Monte Carlo modeling code, supplied by the manufacturer. The geometry of rectangular complex plane for calculating the detection efficiency has been chosen from the ISOCS predefined templates as best matching to the experimental conditions, giving the possibility to include the multilayer design of the entrance door. The energydependent photon detection efficiency (*ε*) of spectrometric system has been finally described with the following function:

ln(ε) = a + b·ln(E) + c·ln(E)<sup>2</sup> + d·ln(E)3 + e·ln(E)4 + f·ln(E)5 , with the fitting parameters of a–f.

The analysis of registered spectra (photopeaks' identification and net areas counting) has been performed using unidentified second differential and nonlinear LSQ fit in Genie™ 2000 software. The sources of gamma radiation (activated nuclides) have been identified on the base of photopeaks' energies, whereas areas under these photopeaks were used for photon flux density (*Φ*) assessment on the basis of Eq. (1), for a defined detector front surface (*SGe*) and life time (*LT*) of each measurement.

## 1

$$
\Phi(E) \quad \left[cm^{-2}s^{-1}\right] = \frac{Peak\\_net\\_area(E)}{\epsilon(E) \cdot S\_{Gk}\left[cm^2\right] \cdot LT\{s\}}
\tag{1}
$$

**165**

*Gamma Radiation in the Vicinity of the Entrance to Linac Radiotherapy Room*

The recommended quantity for the purpose of limitation of ionizing radiation exposure is the effective dose (*Ed*). However, radiation protection and operational quantities are distinguished according to the relevance to radiation health effects and possibility to be measured, respectively. These are presented in reports 26, 60, and 103 of International Commission on Radiological Protection (ICRP) and in reports 66 and 85 of International Commission on Radiation Units and Measurements (ICRU) [47, 49–52]. Moreover, conversion coefficients of radiation fluence to organ absorbed doses as well as equivalent doses and effective dose are also supplied in ICRP 74 and 116 as well as in ICRU 57 reports [53–55]. This enables the risk estima-

Gamma radiation spectrometry system was placed 50 cm in front of linac treatment room entrance door, at the height of 1 m above the floor (see **Figure 4**), at a localization representative for dose rate assessment for the staff waiting for the end

The spectra were registered during emission of 6–18 MV photon beams with a dose rate of 450 MU/min. The gantry angle of 90 or 270° and irradiation field

The detailed characteristics of gamma ray spectra acquired in standard and extended energy range near the linac room door are presented in **Table 1**. The energies of photons in 6 MV beam are too low to trigger nuclear reactions; therefore, induced radionuclides are not observed on the spectrum outside the door in this case, as shown in **Figure 5**. Nevertheless, the increase in low-energy continuous part of the registered spectrum in comparison with natural background radiation indicates that

Spectra registered during high-energy beam emission (10–18 MV), as shown in **Figure 5**, are dominated by two processes: positron creation, since annihilation peak at 511 keV is clearly visible, and neutron capture in hydrogen-rich material inside the door, due to the presence of a peak at 2224.6 keV, which is the neutron-binding energy in deuterium nucleus. The intensity of these processes could be correlated not only with neutron source strength of particular linac working at defined accelerating potential, but also with the amount of hydrogen-rich material used in door construction. The peak at 477.6 keV is due to the presence of boron (mostly in the form of

where 477.6 keV is the deexcitation energy of lithium nucleus. The broadening of this peak has Doppler effect—origin, widely discussed in [56]. Since door construction is not unified and the usage of paraffin/polyethylene (as hydrogen-rich materials) or borax as neutron absorption agents depends on the construction concept, the intensity of the 477.6 keV line should not be directly connected with the therapeutic beam energy (see **Figure 5**: 15 MV vs. 18 MV cases) or even may not occur at all, whereas

The gamma ray spectra are dominated by the abovementioned interactions;

• (n,n'γ) and (n,γ) interactions in germanium crystal of HPGe spectrometer, which proves that neutrons contribute to the door-leakage radiation outside

entrance door, to study the worst scenario of occupational hazard.

part of scattered radiation from therapeutic beam penetrates the door.

borax–sodium tetraborate decahydrate) and is a consequence of 10B(n,α)

hydrogen capture of neutron is present for all linacs studied by us.

however, the minor contributions come from:

the treatment room;

**3.1 Comparison of spectra for various beam energies**

were set to achieve maximal intensity of radiation reaching the

7

Li reaction,

*DOI: http://dx.doi.org/10.5772/intechopen.82726*

tion on the basis of different measuring quantities.

**3. Results and discussion**

of patient irradiation.

size of 40 × 40 cm2

*Gamma Radiation in the Vicinity of the Entrance to Linac Radiotherapy Room DOI: http://dx.doi.org/10.5772/intechopen.82726*

The recommended quantity for the purpose of limitation of ionizing radiation exposure is the effective dose (*Ed*). However, radiation protection and operational quantities are distinguished according to the relevance to radiation health effects and possibility to be measured, respectively. These are presented in reports 26, 60, and 103 of International Commission on Radiological Protection (ICRP) and in reports 66 and 85 of International Commission on Radiation Units and Measurements (ICRU) [47, 49–52]. Moreover, conversion coefficients of radiation fluence to organ absorbed doses as well as equivalent doses and effective dose are also supplied in ICRP 74 and 116 as well as in ICRU 57 reports [53–55]. This enables the risk estimation on the basis of different measuring quantities.
