Feasibility study of optical imaging of the boron-dose distribution by a liquid scintillator in a clinical boron neutron capture therapy field.

BACKGROUND: Evaluation of the boron dose is essential for boron neutron capture therapy (BNCT). Nevertheless, a direct evaluation method for the boron-dose distribution has not yet been established in the clinical BNCT field. To date, even in quality assurance (QA) measurements, the boron dose has been indirectly evaluated from the thermal neutron flux measured using the activation method with gold foil or wire and an assumed boron concentration in the QA procedure. Recently, we successfully conducted optical imaging of the boron-dose distribution using a cooled charge-coupled device (CCD) camera and a boron-added liquid scintillator at the E-3 port facility of the Kyoto University Research Reactor (KUR), which supplies an almost pure thermal neutron beam with very low gamma-ray contamination. However, in a clinical accelerator-based BNCT facility, there is a concern that the boron-dose distribution may not be accurately extracted because the unwanted luminescence intensity, which is irrelevant to the boron dose is expected to increase owing to the contamination of fast neutrons and gamma rays. PURPOSE: The purpose of this research was to study the validity of a newly proposed method using a boron-added liquid scintillator and a cooled CCD camera to directly observe the boron-dose distribution in a clinical accelerator-based BNCT field. METHOD: A liquid scintillator phantom with 10 B was prepared by filling a small quartz glass container with a commercial liquid scintillator and boron-containing material (trimethyl borate); its natural boron concentration was 1 wt%. Luminescence images of the boron-neutron capture reaction were obtained in a water tank at several different depths using a CCD camera. The contribution of background luminescence, mainly due to gamma rays, was removed by subtracting the luminescence images obtained using another sole liquid scintillator phantom (natural boron concentration of 0 wt%) at each corresponding depth, and a depth profile of the boron dose with several discrete points was obtained. The obtained depth profile was compared with that of calculated boron dose, and those of thermal neutron flux which were experimentally measured or calculated using a Monte Carlo code. RESULTS: The depth profile evaluated from the subtracted images indicated reasonable agreement with the calculated boron-dose profile and thermal neutron flux profiles, except for the shallow region. This discrepancy is thought to be due to the contribution of light reflected from the tank wall. The simulation results also demonstrated that the thermal neutron flux would be severely perturbed by the 10 B-containing phantom if a relatively larger container was used to evaluate a wide range of boron-dose distributions in a single shot. This indicates a trade-off between the luminescence intensity of the 10 B-added phantom and its perturbation effect on the thermal neutron flux. CONCLUSIONS: Although a partial discrepancy was observed, the validity of the newly proposed boron-dose evaluation method using liquid-scintillator phantoms with and without 10 B was experimentally confirmed in the neutron field of an accelerator-based clinical BNCT facility. However, this study has some limitations, including the trade-off problem stated above. Therefore, further studies are required to address these limitations.

[Basic Knowledge of Neutron: Generation of Neutrons Accompanied with the High-Energy Photon Therapy].

Photo neutrons are generated from high-energy medical X-ray linacs via photo-nuclear reactions with the materials of target and collimator as well as therapeutic X-rays. Such photo neutrons sometimes make unwanted influences and are not negligible for the aspects of radiation protection and radiation control. In this article, fundamental principle of such photo-neutron generation is briefly explained. The side effects induced by the photo neutrons are summarized. In addition, some techniques of the detection and measurement of photo neutrons are introduced.

First optical observation of 10B-neutron capture reactions using a boron-added liquid scintillator for quality assurance in boron neutron capture therapy.

10B-neutron capture was observed optically using a boron-added liquid scintillator. Trimethyl borate was dissolved in a commercially available liquid scintillator at natural boron concentrations of approximately 1 wt% and 0.25 wt%. The boron-added liquid scintillator was placed in a phantom quartz bottle and irradiated by thermal neutrons (~ 105 n/[cm2 s]) for 150, 300, and 600 s. The luminescence of the liquid scintillator was clearly observed using a cooled charge-coupled device (CCD) camera during irradiation. The luminance value recorded by the CCD camera was proportional to the duration of irradiation by thermal neutrons. The luminescence distribution showed reasonable agreement with that of energy deposition by Li and alpha particles from 10B-neutron capture reactions calculated via Monte Carlo simulations. When trimethyl borate was not dissolved in the liquid scintillator (0 wt% natural boron), no visible luminescence was observed even after 600 s of irradiation. These findings demonstrate that the observed luminance originates from the Li and alpha particles generated by 10B-neutron capture reactions. Consequently, the luminescence distribution is directly related to the boron dose of the liquid scintillator. To the best of our knowledge, direct experimental optical observations of boron dose distribution have not yet been reported. This novel technique will be useful for quality assurance in boron neutron capture therapy (BNCT) because instantaneous neutron irradiation may be sufficient for the observing the intense neutron beam used in clinical BNCT (~ 109 n/[cm2 s]), and quick evaluation of the boron dose distribution is expected to be feasible.

Observation of morphological abnormalities in silkworm pupae after feeding 137CsCl-supplemented diet to evaluate the effects of low dose-rate exposure.

Since the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident, morphological abnormalities in lepidopteran insects, such as shrinkage and/or aberration of wings, have been reported. Butterflies experimentally exposed to radiocesium also show such abnormalities. However, because of a lack of data on absorbed dose and dose-effect relationship, it is unclear whether these abnormalities are caused directly by radiation. We conducted a low dose-rate exposure experiment in silkworms reared from egg to fully developed larvae on a 137CsCl-supplemented artificial diet and estimated the absorbed dose to evaluate morphological abnormalities in pupal wings. We used 137CsCl at 1.3 × 103 Bq/g fresh weight to simulate 137Cs contamination around the FDNPP. Absorbed doses were estimated using a glass rod dosimeter and Monte Carlo particle transport simulation code PHITS. Average external absorbed doses were approximately 0.24 (on diet) and 0.016 mGy/day (near diet); the average internal absorbed dose was approximately 0.82 mGy/day. Pupal wing structure is sensitive to radiation exposure. However, no significant differences were observed in the wing-to-whole body ratio of pupae between the 137CsCl-exposure and control groups. These results suggest that silkworms are insensitive to low dose-rate exposure due to chronic ingestion of high 137Cs at a high concentration.

DEVELOPMENT OF A NEUTRON DOSIMETRY SYSTEM BASED ON DOUBLE SELF-ACTIVATED CSI DETECTORS FOR MEDICAL LINAC ENVIRONMENTS.

In the present study, by using double self-activated CsI detectors, the development of a neutron dosemeter system whose response indicates better agreement with the International Commission on Radiological Protection-74 rem-response was carried out to simply evaluate the neutron dose with high accuracy. The present double neutron dosemeter system, using a slow-neutron dosemeter (thermal to 10 keV) and a fast-neutron dosemeter (above 10 keV), consists of CsI scintillators wrapped with two types of neutron energy filtering materials: polyethylene and B4C silicon rubber. After optimization of each filter thickness, to confirm the validity of our method, the neutron ambient dose equivalents under several operating conditions of medical linear accelerators (Linacs) were evaluated using a Monte Carlo simulation and an experiment with the present dosemeter. From these results, the present dosimetry system has enabled a more accurate neutron dose evaluation than our conventional dosemeter, and the present dosemeter was suitable for the neutron dosimetry for 10 MV Linac environments.

AN EVALUATION OF THE BASIC CHARACTERISTICS OF A PLASTIC SCINTILLATING FIBRE DETECTOR IN CT RADIATION FIELDS.

The ionisation chamber for computed tomography (CT) is an instrument that is most commonly used to measure the computed tomography dose index. However, it has been reported that the 10 cm effective detection length of the ionisation chamber is insufficient due to the extent of the dose distribution outside the chamber. The purpose of this study was to estimate the basic characteristics of a plastic scintillating fibre (PSF) detector with a long detection length of 50 cm in CT radiation fields. The authors investigated position dependence using diagnostic X-ray equipment and dependencies for energy, dose rate and slice thickness using an X-ray CT system. The PSF detector outputs piled up at a count rate of 10 000 counts ms-1 in dose rate dependence study. With calibration, this detector may be useful as a CT dosemeter with a long detection length except for the measurement at high dose rate.

Distribution of residual long-lived radioactivity in the inner concrete walls of a compact medical cyclotron vault room.

OBJECTIVE: Compact medical cyclotrons have been set up to generate the nuclides necessary for positron emission tomography. In accelerator facilities, neutrons activate the concrete used to construct the vault room; this activation increases with the use of an accelerator. The activation causes a substantial radioactive waste management problem when facilities are decommissioned. In the present study, several concrete cores from the walls, ceiling and floor of a compact medical cyclotron vault room were samples 2 years after the termination of operations, and the radioactivity concentrations of radionuclides were estimated. METHODS: Cylindrical concrete cores 5 cm in diameter and 10 cm in length were bored from the concrete wall, ceiling and floor. Core boring was performed at 18 points. The gamma-ray spectrum of each sample was measured using a high-purity germanium detector. The degree of activation of the concrete in the cyclotron vault room was analyzed, and the range and tendency toward activation in the vault room were examined. RESULTS: (60)Co and (152)Eu were identified by gamma-ray spectrometry of the concrete samples. (152)Eu and (60)Co are produced principally from the stable isotopes of europium and cobalt by neutron capture reactions. The radioactivity concentration did not vary much between the surface of the concrete and at a depth of 10 cm. Although the radioactivity concentration near the target was higher than the clearance level for radioactive waste indicated in IAEA RS-G-1.7, the mean radioactivity concentration in the walls and floor was lower than the clearance level. CONCLUSION: The radioactivity concentration of the inner concrete wall of the medical cyclotron vault room was not uniform. The areas exceeding the clearance level were in the vicinity of the target, but most of the building did not exceed the clearance levels.

Applicability of self-activation of an NaI scintillator for measurement of photo-neutrons around a high-energy X-ray radiotherapy machine.

The applicability of the activation of an NaI scintillator for neutron monitoring at a clinical linac was investigated experimentally. Thermal neutron fluence rates are derived by measurement of the I-128 activity generated in an NaI scintillator irradiated by neutrons; ß-rays from I-128 are detected efficiently by the NaI scintillator. In order to verify the validity of this method for neutron measurement, we irradiated an NaI scintillator at a research reactor, and the neutron fluence rate was estimated. The method was then applied to neutron measurement at a 10-MV linac (Varian Clinac 21EX), and the neutron fluence rate was estimated at the isocenter and at 30 cm from the isocenter. When the scintillator was irradiated directly by high-energy X-rays, the production of I-126 was observed due to photo-nuclear reactions, in addition to the generation of I-128 and Na-24. From the results obtained by these measurements, it was found that the neutron measurement by activation of an NaI scintillator has a great advantage in estimates of a low neutron fluence rate by use of a quick measurement following a short-time irradiation. Also, the future application of this method to quasi real-time monitoring of neutrons during patient treatments at a radiotherapy facility is discussed, as well as the method of evaluation of the neutron dose.

[Neutron Dosimetry System Using CR-39 for High-energy X-ray Radiation Therapy].

Neutrons are produced during radiation treatment by megavolt X-ray energies. However, it is difficult to measure neutron dose especially just during the irradiation. Therefore, we have developed a system for measuring neutrons with the solid state track detector CR-39, which is free from the influence of the X-ray beams. The energy spectrum of the neutrons was estimated by a Monte Carlo simulation method, and the estimated neutron dose was corrected by the contribution ratio of each energy. Pit formation rates of CR-39 ranged from 2.3 x 10(-3) to 8.2 x 10(-3) for each detector studied. According to the estimated neutron energy spectrum, the energy values for calibration were 144 keV and 515keV, and the contribution ratios were approximately 40:60 for 10 MV photons and 20:70 for photons over 15 MV. Neutron doses measured in the center of a high-energy X-ray field were 0.045 mSv/Gy for a 10 MV linear accelerator and 0.85 mSv/Gy for a 20 MV linear accelerator. We successfully developed the new neutron dose measurement system using the solid track detector, CR-39. This on-time neutron measurement system allows users to measure neutron doses produced in the radiation treatment room more easily.

[Shape of neutron energy].

Cone-like acryl converters have been used for transforming the energy-distribution information of incident fast neutrons into the spatial-distribution information of recoil protons. The characteristics of neutron-proton conversion have been studied up to around 10MeV by using an imaging plate (IP). A notable and interesting signal enhancement due to recoil protons generated in an acrylic converter was observed on IP images for irradiation with a 252Cf source. A Monte Carlo calculation was carried out in order to understand the spatial distributions of the signal enhancement by recoil protons; these distributions promisingly involve the energy information of incident neutrons in principle. Consequently, it has been revealed that the neutron energy evaluation is surely possible by analyzing the spatial distributions of signal enhancement that is caused by recoil protons.

Feasibility study on using imaging plates to estimate thermal neutron fluence in neutron-gamma mixed fields.

In current radiotherapy, neutrons are produced in a photonuclear reaction when incident photon energy is higher than the threshold. In the present study, a method of discriminating the neutron component was investigated using an imaging plate (IP) in the neutron-gamma-ray mixed field. Two types of IP were used: a conventional IP for beta- and gamma rays, and an IP doped with Gd for detecting neutrons. IPs were irradiated in the mixed field, and the photo-stimulated luminescence (PSL) intensity of the thermal neutron component was discriminated using an expression proposed herein. The PSL intensity of the thermal neutron component was proportional to thermal neutron fluence. When additional irradiation of photons was added to constant neutron irradiation, the PSL intensity of the thermal neutron component was not affected. The uncertainty of PSL intensities was approximately 11.4 %. This method provides a simple and effective means of discriminating the neutron component in a mixed field.

Waveform simulation based on 3D dose distribution for acoustic wave generated by proton beam irradiation.

A pulsed proton beam is capable of generating an acoustic wave when it is absorbed by a medium. This phenomenon suggests that the acoustic waveform produced may well include information on the three-dimensional (3D) dose distribution of the proton beam. We simulated acoustic waveforms by using a transmission model based on the Green function and the 3D dose distribution. There was reasonable agreement between the calculated and measured results. The results obtained confirm that the acoustic waveform includes information on the dose distribution.

Correlation between the respiratory waveform measured using a respiratory sensor and 3D tumor motion in gated radiotherapy.

PURPOSE: The purpose of this study is to investigate the correlation between the respiratory waveform measured using a respiratory sensor and three-dimensional (3D) tumor motion. METHODS AND MATERIALS: A laser displacement sensor (LDS: KEYENCE LB-300) that measures distance using infrared light was used as the respiratory sensor. This was placed such that the focus was in an area around the patient's navel. When the distance from the LDS to the body surface changes as the patient breathes, the displacement is detected as a respiratory waveform. To obtain the 3D tumor motion, a biplane digital radiography unit was used. For the tumor in the lung, liver, and esophagus of 26 patients, the waveform was compared with the 3D tumor motion. The relationship between the respiratory waveform and the 3D tumor motion was analyzed by means of the Fourier transform and a cross-correlation function. RESULTS: The respiratory waveform cycle agreed with that of the cranial-caudal and dorsal-ventral tumor motion. A phase shift observed between the respiratory waveform and the 3D tumor motion was principally in the range 0.0 to 0.3 s, regardless of the organ being measured, which means that the respiratory waveform does not always express the 3D tumor motion with fidelity. For this reason, the standard deviation of the tumor position in the expiration phase, as indicated by the respiratory waveform, was derived, which should be helpful in suggesting the internal margin required in the case of respiratory gated radiotherapy. CONCLUSION: Although obtained from only a few breathing cycles for each patient, the correlation between the respiratory waveform and the 3D tumor motion was evident in this study. If this relationship is analyzed carefully and an internal margin is applied, the accuracy and convenience of respiratory gated radiotherapy could be improved by use of the respiratory sensor.Thus, it is expected that this procedure will come into wider use.