Real-time dosimetry device for neutron capture treatments

Abstract

The present invention relates to a real-time dosimetry device for neutron capture treatments. The device comprises a first module (1) to absorb a flow of dispersed neutrons in a patient which impinge on the device. The first module (1) is made of a lightweight material such as high-density polyethylene, graphite, or a molecular compound, with lithium content. A second module comprises a first segmented gamma radiation detector (2) that is sensitive to radiation interaction positions therein, and a second segmented gamma radiation detector (3), which is also sensitive to the positions, the second gamma radiation detector (3) being detachably coupled to the first gamma radiation detector (2). The device is particularly suitable for real-time dosage monitoring in patients during boron neutron capture treatments (BNCT).

Description

[0001] REAL-TIME DOSIMETRY DEVICE FOR NEUTRON CAPTURE THERAPY FIELD OF THE TECHNIQUE The present invention relates, in general, to dosimetry monitoring devices. In particular, the invention relates to a real-time dosimetry device for neutron capture therapy, particularly boron neutron capture therapy (BNCT). BACKGROUND OF THE INVENTION Due to the proliferation of compact and low-cost accelerators, which can be easily installed in hospitals, neutron capture-based therapies are experiencing a great boom [Porras20]. Clinically, all therapies performed to date have been based on the administration of a drug containing boron-10 ( 10B), typically borophenylalanine (BPA), which selectively accumulates in tumor structures. 10B has a very high neutron capture cross-section (3837 barns for thermal neutrons). After capturing neutrons, 10B transforms into 11 B, which decays into an alpha particle and a nucleus of 7 Li, with a range comparable to cell size. The nucleus of 7 Li is initially in an excited state (94% of the time) and decays to its ground (stable) state by emitting a gamma ray with an energy of 478 keV. The effectiveness of boron neutron capture therapy (BNCT) is based on achieving a sufficient concentration of 10The B dose is high in tumor tissues and minimal in surrounding healthy tissues. On the other hand, the tolerance of healthy tissue to the treatment depends on the secondary dose rate, which is dominated by neutron capture reactions in hydrogen (producing 2.2 MeV gamma radiation) and, to a lesser extent, in nitrogen, as well as by recoil nuclei (mainly protons) produced by elastic neutron collisions during their moderation in tissues. BNCT is a treatment for cases with a difficult prognosis, is particularly interesting for diffuse tumors, and has been clinically studied in various diseases, such as glioblastoma multiforme, meningioma, head and neck cancers, melanomas, lung cancer, breast cancer, etc. [Malouff21]. However, dosimetry in BNCT is a major challenge due to the indirect effect of neutrons on elements present in the body and the uncertain absorption of the boron compound used in clinical practice.Therefore, treatment planning in BNCT is rather unconventional and relies solely on estimates of boron uptake by the tumor, which can be obtained through a prior PET scan of the patient, from which the tumor dose is estimated, and an estimate of the dose to at-risk organs based on blood boron analysis. One possibility for monitoring the concentration of. 10 B in BNCT treatments consists of applying gamma imaging techniques with the 478 keV radiation produced after neutron capture in 10B. This idea has been proposed and studied for over 10 years, using both Compton imaging techniques [Stockhausen12, Lee15, Hou22, Sakai23, Kim23, Ramos23, Nutter24], and techniques based on pinhole cameras [Katabuchi14] and other types of mechanical collimation [Manabe16]. However, to date, a system that functions under clinical conditions and allows for real-time dosimetric data acquisition has not been developed. One of the main reasons this methodology has not been implemented in clinical BNCT treatments is the high neutron background to which the monitor or imaging device is exposed. Interactions with these neutrons within the device itself produce a very high level of gamma radiation noise, hindering the acquisition of an image with sufficient contrast and resolution.In this respect, both at a conceptual level [Domingo16] and experimentally [Lerendegui21,Lerendegui23a-b] the usefulness of neutron absorbers based on light compounds enriched in has been demonstrated. 6 Li to reduce the neutron background induced in the detector itself, especially in Compton-type gamma cameras developed for nuclear astrophysics research [Domingo23]. However, these imaging systems are not applicable in BNCT treatments due to the high counting rates they would be subjected to when placed in close proximity to the patient in a clinical BNCT treatment. It is important to note that to achieve acceptable image resolution in BNCT and distinguish the location of the concentration of 10In the tumor area, relative to the surrounding healthy tissue, the device must be physically located in close proximity to the region of interest. This presents an additional challenge for systems proposed or studied to date, due to the high counting rates induced by the high flux of gamma and neutron radiation in the detectors themselves. To counteract this effect, the monitoring device ends up being positioned relatively far from the object being visualized, which considerably reduces both image resolution and geometric efficiency, as well as real-time monitoring capabilities. Therefore, new real-time dosimetry devices for neutron therapies are needed. References [Babiano19] V. Babiano et al., “First i-TED demonstrator: A Compton imager with Dynamic Electronic Collimation” Nucl. Instr. Meth. A, Volume 953, 11 February 2020, 163228 [Domingo16] C. Domingo-Pardo, Nuclear Instruments and Methods A, Volume 825, 21 July 2016,Pages 78-86 https: / / doi.org / 10.1016 / j.nima.2016.04.002.[Domingo23] C. Domingo-Pardo et al 2023 J. Phys.: Conf. Ser.2586012150 DOI 10.1088 / 1742- 6596 / 2586 / 1 / 012150 [Goorley08] T. Goorley. MCNP Medical Physics Geometry Database. Los Alamos National Laboratory Presentation LA-UR-08-02468. Los Alamos, NM, USA.2008. [Hou22] Z. Hou et al., “Boron concentration prediction from Compton camera image for boron neutron capture therapy based on generative adversarial network”, Applied Radiation and Isotopes 186 (2022) 110302 https: / / doi.org / 10.1016 / j.apradiso.2022.110302 [Katabuchi14] T. Katabuchi et al., “Feasibility study on pinhole camera system for online dosimetry in boron neutron capture therapy” Appl Radiat Isot. 2014 Jun:88:139-42. doi: [Kim23] Kim, Y., Lee, W. Development of a Compton camera based on 3D position-sensitive virtual Frisch-grid CZT detectors. Nucl. Instr. Meth. A, 1053, 168390, 2023 [Lee15] T. Lee et al., “Monitoring the distribution of prompt gamma rays in BNCT using a multiple- scattering Compton camera: A Monte Carlo simulation study” Nuclear Instruments and Methods in Physics Research A 798 (2015) [Lerendegui21] J. Lerendegui-Marco et al., Patent EP4152053A1; ES2877772A1; ES2877772B2; JP2023525136A; US2023288584A1; WO2021229132A1 DEVICE FOR THE SIMULTANEOUS DETECTION, IDENTIFICATION, QUANTIFICATION AND / OR LOCALIZATION OF GAMMA RADIATION AND NEUTRON SOURCES [Lerendegui23a] J. Lerendegui-Marco et al., EPJ Web Conf. Volume 279, 2023 Nuclear Physicsin Astrophysics – X (NPA-X 2022) https: / / doi.org / 10.1051 / epjconf / 202327913001[Lerendegui23b] J. Lerendegui-Marco et al., EPJ Web of Conferences 284, 01028 (2023) https: / / doi.org / 10.1051 / epjconf / 202328401028[Malouff21] T.D. Malouff et al.,”Boron Neutron Capture Therapy: A Review of ClinicalApplications” Front. Oncol.11:601820. https: / / doi.org / 10.3389 / fonc.2021.601820 [Manabe16] Manabe, M., Nakamura, S., Murata, I. Study on measuring device arrangement of array-type CdTe detector for BNCT-SPECT, Rep. Pract. Onc. Radiat.21, 2, pp.102-107, 2016 https: / / doi.org / 10.1016 / j.rpor.2015.04.002 [Nutter24] K. Nutter et al., “A feasibility study using an array of LaBr3(Ce) scintillation detectors as a Compton camera for prompt gamma imaging during BNCT”, Front. Phys., 21 February 2024Sec. Medical Physics and Imaging Volume 12 - 2024 | https: / / doi.org / 10.3389 / fphy.2024.1347929[Porras20] I. Porras et al., “BNCT research activities at the Granada group and the project NeMeSis: Neutrons for medicine and sciences, towards an accelerator-based facility for new BNCT therapies, medical isotope production and other scientific neutron applications”. Appl Radiat Isot.2020 Nov;165:109247. https: / / doi.org / 10.1016 / j.apradiso.2020.109247 [Ramos23] Ramos López, D.; Pugliese, G.M.I.; Iaselli, G.; Amoroso, N.; Gong, C.; Pascali, V.; Altieri, S.; Protti, N. Study of Alternative Imaging Methods for In Vivo Boron Neutron Capture Therapy. Cancers 2023, 15, 3582. https: / / doi.org / 10.3390 / cancers15143582 [Sakai23] Sakai, M., Tamaki, S., Murata, I. et al. Experimental study on Compton camera for boron neutron capture therapy applications. Sci Rep 13, 22883 (2023). https: / / doi.org / 10.1038 / s41598-023-49955-9 [Stockhausen12] L. C. Stockhausen, MPhys Dissertation, “Evaluation of Compton Camera Imaging during Boron Neutron Capture Therapy”, University of Liverpool, May 2012.BRIEF DESCRIPTION OF THE INVENTION The proposed device overcomes the limitations of the prior art by: - ​​Gamma imaging capability that enables real-time monitoring of the spatial distribution of the dose during therapeutic treatments, particularly BNCT. - Ability to withstand the intense neutron field inherent in this type of therapy. - Gamma imaging with spectroscopic capabilities and, therefore, useful for monitoring different dose rates deposited in the patient, mainly the dose rate associated with neutron interactions with boron and neutron interactions with hydrogen. - High counting rate capability, and therefore compatible with the high neutron flux clinical conditions of BNCT, which in turn allows it to be positioned close to the patient and achieve the image resolution necessary for this application. - Small size, modular, scalable, and compatible with the clinical environment.- Relatively lower cost than other known systems, which favors scalability or even the implementation of several elements around the region of interest to minimize systematic errors, facilitate real-time diagnosis, and even apply techniques such as those mentioned in the next point. - It can be rotated around the patient, or several units can be arranged around the patient, to obtain 3D tomographic imaging.Examples of embodiments of the present invention provide, in a first aspect, a real-time dosimetry device for neutron capture treatments, particularly BNCT, comprising a first module configured to absorb a neutron flux scattered by a patient and incident upon the device, wherein the first module is made of a lightweight lithium-containing material; and a second module comprising a first segmented gamma radiation detector sensitive to radiation interaction positions therein and a second segmented gamma radiation detector, with one or more detectors, also sensitive to radiation interaction positions therein. The second segmented gamma radiation detector is detachably coupled to the first segmented gamma radiation detector. According to the present invention, the lightweight material may comprise high-density polyethylene, graphite, or a molecular compound.In some embodiments, the lightweight material is enriched in 6Li. Alternatively, in other embodiments, the lithium content comprises natural lithium. In some embodiments, the first module surrounds the detectors, either wholly or partially. In some embodiments, the first segmented gamma radiation detector and the second segmented gamma radiation detector comprise segmented scintillator crystals of inorganic scintillator such as LaCl3(Ce). In some embodiments, the thickness of the scintillator crystal of the second gamma radiation detector is greater than the thickness of the scintillator crystal of the first gamma radiation detector. In some embodiments, the first gamma radiation detector and / or the second gamma radiation detector is / are sensitive to low-energy neutrons in the thermal range (i.e., 0).0.25 eV) to intermediate energies (e.g., keV) and has / are the ability to discriminate between neutrons and gamma radiation. In some embodiments, one or more of the detectors of the second gamma radiation detector has / are neutron-gamma discrimination capability and is / are sensitive to neutrons with energies between keV and tens of MeV, i.e., high energy. In some embodiments, the second gamma radiation detector comprises a plurality of detectors arranged parallel to each other in different planes. In some embodiments, the device also comprises an auxiliary display module for obtaining anatomical information and providing a spatial reference on which to superimpose the formed dosimetric image. In some embodiments, the device also comprises an acquisition and processing module. It may also comprise a thermal camera associated with the acquisition and processing module.In some implementations, the device can be combined with other sensors, such as RGBD visual and depth image sensors, infrared imaging, ultrasound imaging, etc., to supplement the dosimetric information obtained with anatomical, physiological, or other references. In some implementations, the device can also be complemented with other detection systems, such as neutron dosimeters based on Bonner spheres or equivalent, ionization chambers or equivalent, or other types of radiation detectors. This allows for the acquisition of additional and complementary information during treatment. The proposed device can be scaled up, both in the number of detectors and their size, to obtain a greater amount of information in real time.This can be relevant for minimizing systematic uncertainties, improving dosimetry accuracy, and achieving continuous real-time monitoring. Furthermore, several devices can be positioned around the patient or moved around them, thus enabling the possibility of performing a 3D tomographic reconstruction of the dose distribution within the patient's body. Examples of embodiments of the present invention provide, according to a second aspect, a real-time dosimetry method for neutron capture therapies. The method of the second aspect comprises using the device of the first aspect of the invention, as defined above.Specifically, the proposed method comprises: conducting a preliminary study, for example, Monte Carlo (MC), based on a clinical treatment plan in which the closest or most convenient distance to place the device with respect to a region of interest (ROI) where a tumor to be monitored is located is estimated; positioning the device around the patient, according to the previous study; acquiring data during neutron treatment; processing the data and generating images regularly throughout the treatment; estimating the spatial distribution of 10B concentration in the device's field of view from a measured radiation of 478 keV; estimating a spatial distribution of neutron capture rates in 1H from a measured radiation of 2.2 MeV; contextualize the generated images within an appropriate spatial or anatomical reference; graphically compare the generated images with expected images from a previous study based on a clinical treatment plan; visualize the compared images. Some results obtained with dedicated laboratory measurements and through MC simulation with MCNP6 and Geant4 for neutron fluxes characteristic of BNCT treatments and a Snyder phantom demonstrate that the use of the proposed device, with the introduction of the first module, a neutron absorber containing Lithium, preferably enriched in 6Li, and with a low atomic number Z, with thicknesses between 5 mm and 30 mm (although not necessarily restricted to these values), leads to a reduction and modulation of the neutron flux incident on the gamma camera, while still allowing the Compton technique to be applied for gamma imaging.The Snyder phantom is commonly used as a reference in BNCT studies [Goorley08]. Specifically, for the characteristic neutron field in BNCT treatments, the first module reduces the intensity of the incident neutron flux and modulates it to a specific neutron energy region. In this energy region, two dominant background contributions are avoided: firstly, for Compton chambers based on segmented scintillator crystals with an inorganic scintillator, for example, LaCl3(Ce), thermal capture in the 35Cl of the scintillator is substantially suppressed; and secondly, the main neutron capture resonances are avoided. 35 Cl and 37Low-energy Cl isotopes are the main contributors to the background radiation, primarily due to the lower-energy resonances of 139La (72.3 Ev neutron energy) and 35Cl (398 Ev neutron energy). Without the proposed adaptation of the first module, these two isotopes present in the LaCl3(Ce) scintillating crystal would contribute significantly to the gamma radiation background and significantly hinder the device's operation for the proposed application, as shown later. In the case of semiconductor-based Compton chambers, such as CdZnTe, the lithium-enriched high-density polyethylene absorber, preferably 6Li, also allows for the suppression of neutron capture in the cadmium thermal region, which is essential for operating these types of systems in BNCT.A similar situation occurs with inorganic LaBr3(Ce) scintillators, where the neutron-induced background is significantly reduced by using the first lithium-enriched lithium polyethylene module. In this case, the neutron-induced background and the average neutron capture rate are higher than in LaCl3(Ce) due to the larger neutron capture cross-section of bromine isotopes in the neutron energy region used in BNCT. Finally, the high counting rate expected in a conventional device for this type of treatment is improved in this invention through two modifications or adaptations.On the one hand, the first module, a neutron absorber, already significantly reduces background radiation and counting rates induced directly or indirectly by neutrons within the device itself, especially if the detectors are made of one of the indicated materials: LaCl3(Ce), LaBr3(Ce), or CdZnTe, as shown later. On the other hand, the segmentation or pixelation of the segmented gamma radiation detectors is key to coping with the high gamma radiation fluxes and, consequently, to approaching the region of interest and thus maintaining sufficient image resolution for BNCT applications while achieving a high net counting rate that allows for real-time monitoring. The pixelation of the second module is an alternative to using lead or heavy element shielding, as in [Nutter24], which increases background radiation in the vicinity of the device and would further hinder its clinical implementation.BRIEF DESCRIPTION OF THE DRAWINGS The above and other features and advantages will be more fully understood from the following detailed description of some illustrative, non-limiting embodiments, with reference to the accompanying drawings, in which: Fig. 1 shows a schematic perspective view of the proposed device, with the first module (or neutron-absorbing element or neutron moderator) and the second detection module, consisting of the first gamma radiation detector (2) and the second gamma radiation detector (3). Fig. 2 shows a Monte Carlo (MC) study performed using a Snyder phantom and a neutron beam characteristic of a BNCT treatment. The figure shows the time-coincident energy spectrum deposited between the two Compton chamber crystals, simulated for a system based on LaCl3(Ce) crystals.The gray curve 1 reflects the total spectrum without the first module (1), that is, without the moderator or neutron absorber. The blue curve 2 shows the background radiation contribution due to the neutrons. The combination of absorber (. 6The use of LiPE with LaCl3(Ce) leads to a 17-fold reduction in background level 3, compared to the device without a neutron absorber, improving the signal-to-background ratio by approximately a factor of 5, and the overall counting rate is reduced by two orders of magnitude. The black spectrum 4 represents the final result that would be measured with the device in Fig. 1 based on a LiPE absorber and LaCl3(Ce) crystals. The red spectra 5 and blue 6 show, respectively, the direct gamma contribution from the phantom and the neutron background induced in the device. Fig. 3 shows the simulated deposited energy spectrum for a system based on LaBr3(Ce) crystals. The color representation of the different contributions follows the same criteria as in Fig. 2. In the region of interest (478 keV), the neutron background with the absorber is reduced by a factor of 4, that is, one-quarter of that with LaCl3(Ce).With LaBr3(Ce), the neutron-induced background, which still dominates, is the main contribution to the total spectrum, with the signal-to-background ratio close to 1. Figure 4 shows the energy spectrum deposited in time coincidence between the two crystals of the second module, simulated for a system based on Cs2LiLaBr6(Ce) crystals, commercially known as CLLB. The color representation of the different contributions follows the same criteria as in Figure 2. Figure 5 shows the energy spectrum deposited in time coincidence between the two crystals of the second module, simulated for a system based on CdZnTe crystals. The color representation of the different contributions follows the same criteria as in Figure 2. Figure 6 shows the energy spectrum deposited in time coincidence between the two crystals of the second module, simulated for a system based on HPGe crystals.The graphical representation of colors with the different contributions follows the same criteria as in Fig. 2. Fig. 7 shows the neutron spectrum that would impinge on the second module in a BNCT treatment, as well as the spectrum obtained after passing through a 2cm thick 6LiPE sheet (first module (1) in Fig. 1). The lower figure shows the radiative capture cross-sections in different isotopes present in different materials used in radiation detectors. Fig. 8 shows calculations that demonstrate the effectiveness of 6Li-HDPE for background suppression using a LaCl3(Ce) crystal, in relation to other options for the first module such as PE. natLi-PE and without a neutron absorber. Fig. 9 shows MC calculations of the counting rates obtained with a Snyder phantom irradiated by a neutron flux characteristic of BNCT. The results show the ability to bring the device closer to the region of interest despite the high radiation flux from the phantom, through the segmentation of the scintillator crystals used and the neutron absorber. Fig. 10 illustrates the image resolution for a Compton camera located 200 cm from a 478 keV radiation source and for a source located 10 cm away. This figure shows the importance of being able to support a high counting rate and thus be able to bring the device closer to the region of interest (see Fig. 9). Fig. 11 shows MC calculations that show the counting rate for monolithic sensors and pixelated sensors versus the distance of the device from the surface of the Snyder phantom.Pixelation allows reducing the total counting rate and the first module by two orders of magnitude. 6LiPE by an additional order of magnitude. In total, the combined effect of pixelation and neutron absorption results in a three-order-of-magnitude improvement in counting rate. For a specific maximum counting rate (500 kHz), the combination of these two improvements allows the device to be positioned from a distance of more than 200 cm to as little as 10 cm, which represents the main competitive advantage over other systems described in the prior art using Compton imaging. LaCl3(Ce) and CdZnTe provide equivalent results, with the former being significantly more economical. DETAILED DESCRIPTION OF THE INVENTION AND SOME EXAMPLES OF EMBODIMENT Figure 1 shows an example of the proposed dosimetry device for neutron capture treatments, particularly BNCT.As shown in the figure, the device, which is preferably compact and portable, comprises a first module (1), i.e., a passive neutron-absorbing moderator module, positioned in front of a second module to efficiently absorb part of the neutron flux scattered by the patient and incident upon the device. Specifically, according to the present invention, the first module (1) is designed to surround, at least partially, the second module. The first module (1) is made of a lightweight lithium-containing material. In some embodiments, the lightweight material is high-density polyethylene enriched with the isotope. 6 Li (referred to in this text as 6 LiPE). In other embodiments, the lightweight material comprises a molecular compound enriched in 6 Li just as 6 Li(CH2) nor other lightweight materials with a high neutron absorption coefficient such as natLiH, 6LiH, 6Li2CO3, etc. Note that for BNCT treatments, other commonly used moderators containing boron, such as 10B-(CH2)n, etc., are less suitable than the lightweight materials described above. The first module (1) allows the reduction of two energy components of the neutron field incident on the device: the thermal component is mitigated thanks to the lithium enrichment of the neutron absorber, while the other high-energy component is attenuated by the moderation effect of elastic scattering in the lightweight material itself.For several of the materials studied for gamma radiation detectors, this element allows for the reduction of neutron reactions both in the thermal region (low energies) and in the resolved resonance region (keV energies), thus decreasing the background radiation and the counting rate in the device itself by up to two orders of magnitude. This suppression of background radiation is a fundamental aspect for the actual implementation of these devices in a clinical BNCT treatment. With regard to the second, active detection module, it consists of a first segmented gamma radiation detector (2) and a second segmented gamma radiation detector (3), sensitive to the interaction positions of the measured gamma radiation. In some embodiments, each of these detectors is composed of a segmented crystal of inorganic scintillator such as LaCl3(Ce) or LaBr3(Ce).In this particular case, each detector is coupled to a pixelated photosensor, such as a silicon photomultiplier, a multi-anode photomultiplier, or equivalent. It should be noted that in the case of LaBr3(Ce), the average background is higher due to the greater neutron capture rate in Br, which decreases the signal-to-background ratio. Alternatively, in other embodiments, each detector consists of a CdZnTe or CdTe semiconductor array, achieving similar performance to that of the previous embodiments with an inorganic scintillator. The main drawback is that the complexity and cost of these semiconductor sensors are significantly higher. In a preferred embodiment of the present invention, the segmented scintillators are made of LaCl3(Ce).Generally, the second segmented gamma radiation detector (3) is detachably coupled to the first segmented gamma radiation detector (2). Furthermore, the crystal thickness of the second detector is greater than that of the first detector to optimize the scattering and absorption ratios in both crystals, respectively. Without compromising portability, the second detector can also have a larger cross-sectional area than the first detector to increase the detection efficiency of the system, as described in [Babiano19]. In some embodiments, the segmented gamma radiation detectors (2, 3) can operate as a Compton imaging camera to form one or more images of the incident radiation, primarily 478 keV gamma radiation to monitor neutron reactions in boron and separate images of 2.2 MeV gamma radiation to monitor neutron reactions in hydrogen (water).Therefore, the 478 keV radiation image can be used as an indicator of the B concentration in the tissue (healthy or tumorous) and the neutron absorption rate within it. In an ideal treatment, this rate is expected to be highest in the tumor area and lowest in the surrounding tissues. The 2.2 MeV radiation image, on the other hand, provides an estimate of the secondary dose rate received by the patient (from neutron capture in hydrogen or water), primarily in healthy tissue or independently of the concentration of B. 10 B in the tissue. Furthermore, the 2.2 MeV gamma image can provide a precise and clinically relevant relative anatomical and morphological reference, obtained with the same device and therefore the same methodology, against which to compare the 478 keV image corresponding mainly to the tumor area or the area with the highest absorption rate of 10B. Dedicated neutron beam measurements and MC simulations performed with a first module (1) of high-density polyethylene enriched in 6 Li indicate that the neutron-induced background radiation in LaCl3(Ce) crystals can be reduced by approximately one order of magnitude (see Fig. 2). This unique result stems from a combined effect of the energy modulation induced by the first module (1) based on 6 LiPE, and the energy dependence of the cross-section of 35 The Cl present in LaCl3(Ce) allows the background contribution to be suppressed where the cross-section is highest: in the thermal region and in the region of resolved resonances. Specifically, the first module (1) is especially effective at absorbing thermal neutrons due to the component of 6Li. On the other hand, it also allows for the "smoothing" or reduction of the most energetic component of the neutron field that reaches the second module after being scattered in the patient's tissue during BNCT treatment, thus reducing the contribution to the background neutron captures in the resonances of chlorine isotopes. This absorption effect (in the thermal region) and moderation effect (in the high-energy region) means that, when using an inorganic chlorine-based scintillator, the background contamination in the scintillator itself is particularly low (Fig. 2), compared to other materials such as LaBr3(Ce) (Fig. 3), Cs2LiLaBr6(Ce), commercially known as CLLB and referred to as such in this text (Fig. 4), etc., which have a higher neutron capture cross-section in this energy range. Regarding the neutron-induced background, CdZnTe (Fig. 5) is a comparable alternative in performance to LaCl3(Ce).However, its implementation in BNCT may be limited by the higher cost and complexity of these types of detectors. For LaBr3(Ce) crystals (Fig. 3) or CLLB (Fig. 4), the main limitation is the high neutron capture cross-section of bromine isotopes and the corresponding increase in background noise within the device itself. In the case of HPGe semiconductor detectors (Fig. 6), the spectroscopic and sensitivity performance is excellent, but their main limitation is their slow time response and their inability to operate at the high counting rates required in this application, in addition to their limited portability and difficult compatibility with the clinical environment. In any case, it is important to emphasize that the present invention is not limited to any specific detector material, and any of the indicated materials can be used, particularly LaCl3(Ce), LaBr3(Ce), CdTe, or CdZnTe. The importance of combining the first module (1), as...6LiPE, with a suitable detection material such as LaCl3(Ce), is shown in Fig. 7. The upper panel of this figure shows the neutron flux that would impinge on the device in a BNCT treatment with and without a neutron absorber. The lower panel shows the main isotopes that, depending on the material chosen for the gamma radiation detectors (2, 3), can contribute to the background radiation induced by neutron capture reactions. As described above, the convolution of the neutron spectrum after passing through the first module (1) with the capture spectrum is what results in a higher or lower background rate in the device. The particularly favorable effect for BNCT of the first module (1) based on 6LIPE is demonstrated in Fig. 8. Using a detection system based on LaCl3(Ce) crystals as a reference, Fig.Figure 8 shows the reduction of the background by means of a common polyethylene absorbent, high-density polyethylene enriched with natural lithium and. 6 LiPE with respect to the situation where no first module (1) would be used, i.e., without a neutron absorber. In summary, the combination of LaCl3(Ce) and 6LiPE allows for an order of magnitude improvement in background noise, thus improving the signal-to-background ratio and decreasing the overall counting rate, which are two of the main limitations of the systems described to date in the state of the art for BNCT. Through dedicated laboratory measurements and MC simulations, it has been verified that thicknesses of the lightweight 6LiPE material between 10 mm and 30 mm offer optimal results with regard to the signal-to-background ratio, although other thinner or thicker thicknesses could still be useful in specific clinical cases or facilities, depending on the specific neutron spectrum used in the BNCT treatment. The first module (1) can completely enclose the second module (as illustrated in Fig. 1), or it can simply be a sheet of a certain thickness. The first option would have additional background suppression advantages, but would naturally be more expensive.The proposed device, in some implementation examples, can be complemented with a data acquisition and processing system, as well as the equipment necessary for data analysis and visual representation or transfer to other equipment (not shown in Fig. 1). As mentioned previously, another problem with state-of-the-art monitoring systems is the high counting rate they would encounter in a real clinical setting or treatment. The counting rate obtained with a system based on monolithic crystals is reduced to acceptable values ​​for this application in BNCT by combining the first 6LiPE module (1), which reduces the number of detected events, with crystal segmentation that allows the flux to be distributed individually among a large number of elements (64 per detector in Fig. 1). These results are shown in Fig.Figure 9, where the 500 kHz limit is arbitrarily indicated as an acceptable counting rate in a real-world application. Segmenting the second module allows for a very significant improvement for this application, as it is key to reducing the counting rate per reading channel by an order of magnitude, thus enabling the gamma camera to be brought closer to the region of interest (ROI). This directly impacts the image resolution and the monitoring capabilities. This is shown in Figure 10. The image of the gamma radiation source can be obtained using the Compton technique, in which a Compton cone can be calculated event by event from the energies measured in the gamma radiation detectors (2, 3) and the interaction positions of the gamma radiation measured in them. The aperture of this cone is given by the Compton scattering law for radiation: where Eγ,i and Eγ,f are the energies of the incident gamma ray and the scattered gamma ray, respectively, h is Planck's constant, mec2 represents the mass of the electron (511 keV), and θ corresponds to the Compton scattering angle formed between the direction of the incident gamma ray and the scattered gamma ray. For a sufficiently thick absorbing detector, it can be assumed that E γ,f = E2, where E2 is the energy measured with the gamma radiation detector. If the energy of the incident radiation is unknown, it can be assumed, as a good approximation, for a wide range of energies, that E γ,i= E1 + E2, where E1 is the energy measured in the first detector (2). This approximation is acceptable for the gamma radiation to be monitored at 478 keV and 2.2 MeV. Image reconstruction of the gamma radiation source is possible using backprojection methods, maximum-likelihood expectation maximization statistical methods, and other similar techniques widely described in the scientific literature. In summary, the combination of the first module (1) of lightweight material enriched with lithium content, for example, high-density polyethylene enriched with the isotope 6LiO6LiPE, segmentation of the radiation-sensitive volume, and the selection of a suitable material in the gamma radiation detectors, for example, LaCl3(Ce), allow for a reduction of several orders of magnitude in the overall counting rate compared to conventional systems, thus enabling high image resolution and real-time image reconstruction, as required for dosimetry in BNCT treatments. These characteristics represent the main competitive advantage over the systems described to date. The present invention also relates to a real-time dosimetry method for neutron capture treatments, which allows for the monitoring of the neutron reaction rate with Boron (478 keV) and with Hydrogen (2.2 MeV). The method in this exemplary embodiment comprises the following steps: 1.1. Preliminary MC Study: A preliminary MC study is performed based on a clinical treatment plan, estimating the optimal distance for placing the device relative to the region of interest where the tumor to be monitored is located. 2. Device Positioning: The device is placed around the patient, according to the results obtained in the preliminary study. 3. Data Acquisition: Data acquisition begins once neutron therapy has started. 4. Visible Image Acquisition: Images are recorded in the visible spectrum using RGB or RGB-D cameras, with the possibility of supplementing them with machine vision techniques. 5. Compton Image Processing (478 keV): Compton images are processed corresponding to 478 keV radiation to monitor neutron reactions with boron. 6. Compton Image Processing (2.2 MeV): Compton images are processed corresponding to 2.2 MeV radiation.2 MeV for monitoring neutron reactions with hydrogen. 7. Anatomical image superimposition: Superimposing anatomical images obtained by visible and CT / MRI techniques with the 478 keV (boron) Compton images and the 2.2 MeV (hydrogen) images. 8. Comparison of results: Comparing the results obtained with the previous MC study to evaluate the accuracy of the treatment and adjust parameters if necessary. The scope of the present invention is defined in the appended claims.

Claims

CLAIMS 1. A real-time dosimetry device for neutron capture treatments, comprising: - a first module (1), configured to absorb a neutron flux scattered by a patient and incident on the device, wherein the first module (1) is made of a lightweight material containing lithium, wherein the lightweight material includes high-density polyethylene, graphite, or a molecular compound; and - a second module, comprising: a first segmented gamma radiation detector (2) sensitive to radiation interaction positions therein; and a second segmented gamma radiation detector (3) sensitive to radiation interaction positions therein, wherein the second gamma radiation detector (3) includes one or more detectors and is detachably coupled to the first gamma radiation detector (2).

2. The device of claim 1, wherein the lithium content is enriched en 6Li.

3. The device of claim 1, wherein the lithium content comprises natural lithium.

4. The device of any one of the preceding claims, wherein the first gamma radiation detector (2) and the second gamma radiation detector (3) comprise segmented scintillator crystals of inorganic scintillator or semiconductor material.

5. The device of claim 4, wherein the inorganic scintillator comprises LaCl3(Ce) or LaBr3(Ce).

6. The device of claim 4, wherein the semiconductor material comprises CdZnTe or CdTe.

7. The device of claim 4, 5, or 6, wherein the thickness of the scintillator crystal of the second gamma radiation detector (3) is greater than the thickness of the scintillator crystal of the first gamma radiation detector (2).

8. The device of any one of the preceding claims, wherein the first gamma radiation detector (2) and / or the second gamma radiation detector (3) is / are sensitive to low-energy neutrons in the thermal range of 0.025 eV to intermediate energies keV and has / have the ability to discriminate between neutrons and gamma radiation.

9. The device of any one of the preceding claims, wherein one or more of the detectors of the second gamma radiation detector (3) has / have neutron-gamma discrimination capability and is / are sensitive to neutrons of energy between keV and tens of MeV.

10. The device of any one of the preceding claims, wherein the second gamma radiation detector (3) comprises a plurality of detectors arranged parallel to each other in different planes.

11. The device of any one of the preceding claims, wherein the first gamma radiation detector (2) and the second gamma radiation detector (3) are configured to operate as a Compton imaging camera to form one or more images of the incident gamma radiation. 12.The device of any one of the preceding claims, further comprising an auxiliary display module configured to obtain anatomical information and provide a spatial reference on which to superimpose the formed dosimetric image.

13. The device of any one of the preceding claims, further comprising an acquisition and processing module, and a thermal camera associated with the acquisition and processing module.

14. The device of any one of the preceding claims, wherein the molecular compound comprises 6LiH, 6Li2CO3, 6Li(CH2)n.

15. The device of any one of the preceding claims, wherein the first module (1) is adapted to surround, totally or partially, the first gamma radiation detector (2) and the second gamma radiation detector (3).

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