RU2709682C1 - Method for determining absorbed dose from thermal neutrons in boron-neutron capture therapy of malignant tumors - Google Patents

Abstract

FIELD: medicine.

SUBSTANCE: invention refers to nuclear medicine, namely to neurooncology, and can be used to determine absorbed dose from thermal neutrons in boron-neutron capture therapy of malignant tumors. Patient is administered the targeted boron delivery preparation. Epithermal neutrons are flown. Measuring γ-spectrometer radiation intensity γ-quanta of atomic nuclei activated with epithermal neutrons. Reagents with atomic nuclei activated by epithermal neutrons are sodium, chlorine, potassium and manganese contained in the human body. Absorbed dose of thermal neutrons D is determined from relationship D = D_calc × Y_meas / Y_calc, where D_calc – absorbed dose from thermal neutrons, obtained by calculation, Y_meas is radiation intensity γ-quantities of atomic nuclei activated with epithermal neutrons, measured γ-spectrometer, Y_calc is radiation intensity γ-quanta of atomic nuclei activated with epithermal neutrons, obtained by calculation.

EFFECT: method provides determining the absorbed dose from thermal neutrons.

3 cl, 1 tbl, 1 dwg

Description

The invention relates to nuclear medicine, in particular to neurooncology, and can be applied when carrying out boron-neutron capture therapy (BNCT) of malignant tumors to determine the absorbed dose.

The concept of neutron capture therapy in oncology was proposed in 1936, four years after the discovery of the neutron. Its physical principle is as follows. A solution containing the stable isotope boron-10 is introduced into the blood of a person, and after a while boron is sorbed mainly in the tumor cells. Then the tumor is irradiated with a stream of epithermal (with energies from 0.5 eV to 10 keV) neutrons. As a result of neutron absorption by the stable ¹⁰ V isotope, a nuclear reaction occurs, and the resulting energetic α-particle and ⁷ Li ion are rapidly inhibited along the length of the cell size and release energy of ~ 2.3 MeV within the very cell that contained the boron nucleus, which leads to her defeat. Thus, boron neutron capture therapy (BNCT) allows for the selective defeat of malignant tumor cells.

The feasibility of developing neutron capture therapy technology is due to its focus on the treatment of such types of malignant tumors that are practically not amenable to any other methods, such as brain glioblastomas and melanoma metastases.

In the early 1950s, Dr. Sweet demonstrated for the first time that certain boron compounds provide a higher concentration of boron in cancer cells compared to healthy ones [W. Sweet The uses of nuclear disintegration in the diagnosis and treatment of brain tumor. N. Engl. J. Med. 245 (1951) 875-878; W. Sweet, M. Javid. The possible use of slow neutrons plus boron-10 in the therapy of intracranial tumors. Trans. Am. Neurol. Assoc. 76 (1951) 60-63]. Since the mid-1950s, the methodology has been tested with varying success at nuclear reactors in a number of countries. The main achievements were associated with the name of the Japanese neurosurgeon Hiroshi Hatanaka. Hatanaka began to use the intra-arterial administration of ¹⁰ V isotope-enriched sodium borocaptate (BSH, Na ₂ ¹⁰ B ₁₂ H ₁₁ SH), synthesized by Dr. Soloway [A. Soloway, N. Hatanaka, M. Davis. Penetration of brain and brain tumor. VII. Tumor binding sulfhydryl boron compounds. J. Med. Chem. 10 (1967) 714-717], performed open irradiation of the tumor after surgical treatment and achieved impressive results - the 5-year survival rate was 58% for a group of patients with malignant gliomas of grade 3 and 4 [Hatanaka N. Clinical results of boron neutron capture therapy . Basic Life Sci. 54 (1990) 15-21]. In 1987, Mishima treated superficial malignant melanoma using ¹⁰ V enriched boronylalanine in the optically isomeric form L (BPA, (HO) ₂ ¹⁰ BC ₆ H ₄ -CH ₂ CH (NH ₂ ) -CO ₂ H) [Y. Mishima, et al. Selective thermal neutron capture therapy and diagnosis of malignant melanoma: from basic studies to first clinical treatment. Basic Life Sci. 50- (1989) 251-260]. The use of these pharmaceuticals allows you to create a concentration of ¹⁰ V isotope in tumor tissue up to 40 μg / g, which is 3.5 times more than in healthy tissue. Such concentration and contrast make it possible to make the contribution of background radiation reasonably small, and indeed provide the possibility of selective damage to tumor cells.

One of the main problems requiring a solution for introducing boron-neutron capture therapy into clinical practice is the determination or measurement of the absorbed dose, both the dose received by the tumor cells and the healthy ones. In contrast to γ-therapy, proton or ion, in BNCT, in addition to determining the characteristics of ionizing radiation, it is necessary to know the spatial distribution of boron in the patient's body.

The following processes contribute to the absorbed dose.

Firstly, the process of neutron absorption by boron, as a result of which 2.79 MeV is released, in 6.1% of cases the energy is distributed only between lithium nuclei and the α-particle, in 93.9% of cases the lithium nucleus flies out in an excited state and emits γ-ray energy of 0.48 MeV. This dose component is called the boron dose.

Secondly, neutron capture by hydrogen nuclei, which leads to the formation of deuterium and the emission of a γ quantum with an energy of 2.2 MeV, neutron capture by nitrogen nuclei, which leads to the formation of a ¹⁴ C nucleus and a recoil proton with the release of energy of 580 keV, neutron capture by chlorine nuclei, leading to the emission of gamma rays with energies up to 8.85 MeV. This dose component is called the dose from thermal neutrons.

Thirdly, the appearance of recoil nuclei in the elastic scattering of neutrons, mainly fast, from the nuclei of matter, mainly hydrogen. This dose component is called the fast neutron dose.

Fourth, the flux of gamma rays from the target site, including the target and the neutron beam formation system. This dose component is called the dose from gamma rays. Quite often, the dose from γ-quanta is understood not only as the dose from an external source (target site), but also from an internal (irradiated object), when γ-quanta are formed as a result of absorption of a neutron by hydrogen or chlorine. In this case, the dose from γ-quanta is understood only as the dose from an external source of γ-radiation, and the dose emitted by γ-quanta due to absorption of the neutron by hydrogen and chlorine refers to the dose from thermal neutrons.

During therapy, each of these components contributes to the absorbed dose, and some of them make a determining contribution. So, the boron dose is the determining dose when irradiating tumor cells in which boron is accumulated, for example, glioblastomas. The dose from fast neutrons is significant for the skin, through the surface of which the neutron flux is directed to the tumor. The dose from thermal neutrons determines the dose received by cells of healthy organs, for example, brain cells in the treatment of glioblastoma.

Measurement of dose components is an important task when conducting BNCT.

Currently, most methods of measuring the absorbed dose in tumor tissues are associated with determining the concentration of boron in the tumor and surrounding tissues and with subsequent numerical calculations of neutron transfer and γ-radiation in the phantom with a given spatial distribution of boron concentration. As a rule, the collection of tumor cells is used to measure boron concentration. The boron concentration in the samples can be determined using several methods: atomic emission spectroscopy [A. Wittig, et al. Boron analysis and boron imaging in biological materials for boron neutron capture therapy (BNCT). Crit. Rev. Oncol. Hematol. 68 (1) (2008) 66-90]; high-resolution alpha autoradiography [WS Kiger III, et al. Boron microquantification in oral muscosa and skin following administration of a neutron capture therapy agent. Radiat. Prot. Dosimetry 99 (1-4) (2002) 409-412]; neutron capture radiography [R. Pugliesi, M. Pereira. Study of the neutron radiography characteristics for the solid state nuclear track detector makrofol-de. NIM A 484 (2002) 613-618]; laser secondary neutron mass spectrometry [P. Binns, et al. An international dosimetry exchange for boron neutron capture therapy, part 1: absorbed dose measurements. Med. Phys. 32 (2005) 3729-3736]; electron microscopy [Y. Zhu, R. Egerton, M. Malac. Concentration limits for the measurement of boron by electron energy loss spectroscopy and electron-spectroscopic imaging. Ultramicroscopy 87 (2001) 135-145]; using fluorescent staining of the contents of tumor agents; magnetic resonance imaging; positron emission tomography [K. Ishiwata, et al. A unique in vivo assessment of 4- [ ¹⁰ B] borono-L-phenylalanine in tumor tissues for boron neutron capture therapy of malignant melanomas using positron emission tomography and 4-borono-2- [ ¹⁸ F] fluoro-L-phenylalanine. Melanoma Res. 2 (1992) 171-179].

All the methods listed above allow only the boron distribution to be determined. To determine the absorbed dose, it is necessary to calculate the transfer of neutrons and γ-radiation, which is carried out by the MCNP program developed at the Los Alamos National Laboratory (USA) [X-5 Monte Carlo Team (2003) MCNP - a general Monte Carlo N-particle transport Code , Version 5. Los Alamos National Laboratory, LA-UR-03-1987], or the PRISMA program developed at the Federal Nuclear Center Snezhinsk (Russia) [M. Arnautova, et al. Monte-Carlo simulation in nuclear geophysics: Comparison of the PRIZMA Monte Carlo program and benchmark experiments. Nuclear Geophysics 7 (1993) 407-418], using the database of ENDF / B sections of different versions and the reference constant [S.N. Abramovich et al. Nuclear physics constants of thermonuclear fusion. Moscow, Central Research Institute of Atominform, 1989]. This procedure is widely used in BNCT. When determining the absorbed dose, all four dose components are calculated: the boron dose, the dose from thermal neutrons, the dose from fast neutrons and the dose from γ-quanta.

A method that directly measures the boron dose in a tumor is γ-spectroscopy, based on the detection of a 478-keV γ-ray emitted during the instantaneous decay of a boron nucleus after absorption of a neutron [T. Kobayashi, K. Kanda. Microanalysis system of ppm order B-10 concentrations in tissue for neutron capture therapy by prompt gamma-ray spectrometry. Nucl. Instrum. Methods Phys. Res. 204 (1983) 525-531].

However, the application of this method for BNCT seems unlikely. There are several reasons. First, all γ-ray recording equipment over a large number of channels should be located and work near the patient in the process of irradiation with a neutron flux. Secondly, the neutron beam formation system, consisting of a moderator, a reflector and an absorber, is close to the patient and will not allow recording equipment to be placed in at least half of the space. Thirdly, the detectors will register not only useful signals from the decay of boron nuclei, but also spurious signals from γ-quanta and neutrons.

The second method, which directly allows you to measure the boron dose in a tumor, is the patented method [S.Yu. Taskaev, A.A. The fence A method for measuring the absorbed dose in boron-neutron capture therapy of malignant tumors. Patent for invention No. 2606337 dated 01/10/2017]. It is implemented as follows. Before irradiation, in accordance with the BNCT procedure, the patient is given a boron targeted delivery drug, additionally labeled with a stable atomic nucleus, characterized by a large cross-section of radiation capture of neutrons, best of all a gold nucleus, possibly indium or silver. In the process of irradiation with epithermal neutrons, in addition to the process of neutron absorption by boron, which leads to instant decay of the excited boron nucleus and the release of energy (accumulation of the absorbed dose), the process of radiation capture of neutrons by marked atomic nuclei will occur, leading to the appearance and accumulation of radioactive nuclei. If the half-life of the resulting radioactive nuclei is greater than or comparable with the characteristic time of neutron irradiation (about 1 hour), then the induced activity can be measured with a γ-spectrometer located outside the irradiation room and in the absence of neutron generation.

Thus, the measurement of the ratio of the concentrations of boron and target nuclei for radiation capture of neutrons and the measurement of the induced activity after irradiation make it possible to restore the spatial distribution of the absorbed dose.

The dose in healthy tissue is determined similarly: by performing numerical calculations of neutron transfer and γ-radiation in a phantom with a given spatial distribution of boron concentration. Since the main contribution to the dose in healthy tissues, with the exception of the skin, is made by the capture of neutrons by the nuclei of hydrogen, nitrogen and chlorine, the uncertainty with the concentration of boron is not as significant as in the case of the dose in the tumor. In other words, the results of calculating the dose in healthy tissues are more reliable than tumors, including because a great deal of experience has been gained in the treatment of x-ray and γ-radiation. At the same time, a direct method for measuring dose in healthy tissues or a component of doses is extremely relevant.

When conducting BNCT in nuclear reactors, a method is used that indirectly gives an idea of the dose in healthy tissues. For example, an X-ray film is used at the KURRI reactor at the Institute of Integrated Radiation and Nuclear Sciences (previously until 03/01/2018: Institute for Reactor Research) at Kyoto University (Japan). The patient is rolled up to the vertical channel of the neutron beam on a trolley on which a collimator is mounted. During irradiation, a collimator is placed between the reactor and the patient. After irradiation, the cart with the patient and the collimator moves away from the reactor and, while the patient is in the same position as during the irradiation, an x-ray film is placed on the collimator from the side of the reactor, which is illuminated by the radiation of the nuclei of neutron-activated substances. The degree of exposure of the x-ray film provides information on the degree of activation of the tissue subjected to neutron irradiation, and qualitatively on the dose, the knowledge of which is considered useful in the analysis of therapy.

The disadvantage of this method is that it does not allow to determine the spatial distribution of the dose from thermal neutrons and to establish the maximum dose received by healthy tissue, which is critically important when planning therapy.

Closest to the proposed method can be considered a method of measuring the absorbed dose in a tumor, protected by a patent for the invention [S.Yu. Taskaev, A.A. The fence A method for measuring the absorbed dose in boron-neutron capture therapy of malignant tumors. Patent for invention No. 2606337 dated 01/10/2017].

This method of measuring the absorbed dose in a tumor is implemented as follows. Before irradiation, in accordance with the BNCT procedure, the patient is given a targeted delivery drug of boron marked with a stable atomic nucleus, characterized by a large cross section for radiation capture of neutrons - a gold or indium nucleus. In the process of irradiation with epithermal neutrons, in addition to the process of neutron absorption by boron, which leads to instant decay of the excited boron nucleus and the release of energy (accumulation of the absorbed dose), the process of radiation capture of neutrons by marked atomic nuclei will occur, leading to the appearance and accumulation of radioactive nuclei. After BNCT is carried out using a γ-spectrometer located in a separate room, the spatial distribution of the radiation intensity of γ-rays with an energy of 411 keV (in the case of gold) is measured and, taking into account the ratio of the activation of gold to the neutron absorption intensity obtained by calculation, the spatial distribution is restored absorbed dose.

The disadvantage of this method can be considered that it allows you to measure the boron dose, and not the dose from thermal neutrons.

The present invention aims to create a method that allows you to determine the absorbed dose from thermal neutrons during BNCT. This component of the dose from thermal neutrons makes a decisive contribution to the total absorbed dose received by healthy tissue.

The technical result of the invention is the determination of the absorbed dose from thermal neutrons during BNCT.

The essence of the proposed method for determining the absorbed dose from thermal neutrons during BNCT is as follows:

Prior to BNCT, the patient is given a drug of targeted delivery of boron in order to accumulate boron in a sufficient concentration in the tumor. Conducting BNCT consists in irradiating the patient with a stream of epithermal neutrons, despite the fact that the tumor contains boron in a concentration sufficient for the therapeutic effect.

In the process of irradiation with epithermal neutrons, neutron capture by hydrogen nuclei leads to the formation of deuterium and to the emission of a γ-ray energy of 2.2 MeV, neutron capture by nitrogen nuclei, which leads to the formation of a ¹⁴ C nucleus and a recoil proton with the release of energy of 580 keV, neutron capture by nuclei chlorine, leading to the emission of γ-quanta with energies up to 8.85 MeV. Using the method of numerical simulation of neutron transfer and γ-radiation using MCNP or PRISMA programs, the absorbed dose from thermal neutrons is calculated; we denote this value D _calc .

In the process of irradiation with epithermal neutrons, in addition to the absorption of neutrons by the nuclei of hydrogen, nitrogen and chlorine with the release of energy, there will be radiation capture of neutrons by stable atomic nuclei, leading to the appearance and accumulation of radioactive nuclei. Using the method of numerical simulation of neutron transfer and γ-radiation using MCNP or PRISMA programs, the radiation intensity of γ-quanta of atomic nuclei activated by epithermal neutrons is calculated; denote this value Y _calc .

After conducting therapy (irradiation with epithermal neutrons) using a γ-spectrometer located in a separate room and in the absence of neutron generation, measure the radiation intensity of γ-quanta of atomic nuclei activated by epithermal neutrons; we denote this value Y _meas .

The absorbed dose from thermal neutrons is determined from the ratio

Stable target nuclei for radiation capture of neutrons in the human body are such as nuclei, sodium, chlorine, potassium and manganese. The table shows the characteristics of these stable nuclei and radioactive nuclei formed by the absorption of neutrons.

The cross section for radiation capture of neutrons by sodium, chlorine, potassium, and manganese nuclei is small, but due to the high content of these elements in the human body, the radiation intensity of γ-quanta of atomic nuclei activated by epithermal neutrons is sufficient for measurement by a γ-spectrometer. The half-life of radioactive nuclei from half an hour to 15 hours is acceptable for accurate measurement of the radiation intensity of gamma rays. The atomic nuclei of sodium, chlorine, potassium, and manganese activated by epithermal neutrons emit gamma rays with energies of 0.847, 1.369, 1.524, and 1.640 MeV.

Thus, the proposed method for determining the absorbed dose from thermal neutrons during BNCT can be implemented in practice as follows. The patient is given a drug for targeted delivery of boron, irradiated with a stream of epithermal neutrons, the absorbed dose from thermal neutrons (D _calc ) is calculated, the radiation intensity of γ-quanta of atomic nuclei (Y _calc ) activated by epithermal neutrons is _calculated , then, after irradiation with a γ-spectrometer, the intensity is measured radiation of γ-quanta of atomic nuclei of sodium, chlorine, potassium and manganese, (Y _ISM ) activated by epithermal neutrons, and the absorbed dose from thermal neutrons is determined from the ratio

the γ-spectrometer measures the radiation intensity of γ-quanta with energies of 0.847, 1.369, 1.524 and 1.640 MeV. emitted by ²⁴ Na, ⁵⁴ Mn, ³⁸ Cl and ⁴² K,

The feasibility of the proposed dose measurement method is illustrated in FIG. 1. In FIG. Figure 1 shows the dependence of the radiation intensity Y of a laboratory mouse after its irradiation with epithermal neutrons on the energy of γ-quanta E. The emission spectrum was measured by a SEG-1KP-IFTP γ-spectrometer based on a semiconductor detector made of especially pure germanium. The emission lines of ²⁴ Na, ⁵⁴ Mn, ³⁸ Cl, and ⁴² K are clearly visible on the spectrum, due to neutron capture by the nuclei of sodium, chlorine, potassium, and manganese present in the animal.

Claims (5)

1. The method of determining the absorbed dose from thermal neutrons during boron-neutron capture therapy of malignant tumors, including the introduction of a patient drug targeted delivery of boron, irradiation with a stream of epithermal neutrons and measuring a γ-spectrometer of the radiation intensity of γ-quanta of atomic nuclei activated by epithermal neutrons, characterized in that As reagents with atomic nuclei activated by epithermal neutrons, sodium, chlorine, potassium and manganese contained in the human body are used, while th dose of thermal neutrons is determined from the ratio D

where D _calc is the absorbed dose from thermal neutrons obtained by calculation, Y _ISM is the radiation intensity of γ-quanta of atomic nuclei activated by epithermal neutrons, measured by a γ-spectrometer, Y _calc is the radiation intensity of γ-quanta of atomic nuclei activated by epithermal neutrons, obtained by calculation. 2. The method according to p. 1, characterized in that the γ-spectrometer measures the radiation intensity of γ-quanta with energies of 0.847, 1.369, 1.524 and 1.640 MeV. 3. The method according to p. 1, characterized in that the γ-spectrometer measures the radiation intensity of γ-quanta for up to 15 hours after irradiation.

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