Radiation Measurements最新文献

The changes in the neutron spectrum produced by the AmBe neutron source moderated in the polyethylene sphere need to be characterized. A passive single-moderator neutron spectrometer with indium foil activation neutron detector has been developed to characterize the neutron spectrum of an AmBe source inside a polyethylene sphere. The detector response function was calculated using the MCNPX 2.7 program with IRDFF-II nuclear data library. Measurements were conducted by placing the Single-Cylindrical Neutron Spectrometer (SCNS) on the outer surface of the sphere for 5 h. The ^116mIn activity due to neutron activation in each indium foil was measured using a gamma spectrometer with an HPGe detector. By identifying the count value at the peak energy of 1294 KeV and considering an HPGe detector efficiency of 3.2% at the foil position 1 mm above the detector surface, the activity of ^116mIn was obtained. The activity value of ^116mIn from each indium foil was compared with the MCNPX simulation results. The neutron spectrum was unfolded using the UMG 3.3 program with activity data input of ^116mIn for each foil and detector response. A neutron spectrum was obtained with a total neutron flux of 634 ± 60 n/cm²·s, consisting of 25% thermal neutrons, 16% epithermal neutrons, and 61% fast neutrons. When compared with the simulation results, the total neutron flux in the spectrum produced by SCNS-In showed only a small difference of 1%. Based on these neutron spectrum measurements, it was determined that placing the AmBe neutron source inside a 15″ diameter PE-sphere will reduce the fast neutron flux by 78%.

A nuclear criticality results in the emission of both neutron and gamma radiation and can produce doses to personnel near the event that exceed 0.1 Gy (10 rad). The primary purpose of nuclear accident dosimetry is to rapidly identify affected personnel in need of prompt medical treatment and to reassure personnel who have been only minimally exposed. While accurate dosimetry is desired, it must be recognized that dose determinations made from whole-body dosimeters or simple triage methods are very rough estimates and contain significant uncertainties. Even when accounting for factors like varying neutron energy spectra, mean photon energies, body orientation within the radiation field, and transient effects on dosimeter response, etc., the end value is a dosimetric quantity defined for very specific radiological conditions and determined within a simple phantom usually at a single depth. Of more importance is the biological response to the radiation, which will vary by person and can be affected by the individual's radiation sensitivity, age, gender, mass, and underlying health conditions. The overall biological, person-specific response to a given dose cannot be precisely determined except by patient symptom observation and individual biological dosimetry (e.g. chromosome analysis, lymphocyte ratios, etc.). This work describes and discusses the criticality accident dosimetry program at the Y-12 National Security Complex, a United States Department of Energy National Nuclear Security Administration facility. The primary goals of the Y-12 accident dosimetry program are, among others, the rapid identification of significantly exposed persons, prompt routing of exposed workers for medical evaluation and treatment, and the ultimate processing of dosimeters to assign doses to personnel.

This study investigates pediatric Computed Tomography (CT) dosimetry, specifically in the thoracic and head regions, using thermoluminescent dosimeters (TLD). The investigation, conducted at the Imaging Diagnostic Center of the Dom Vicente Scherer Hospital, involved two CT machines, the GE Optima 540 and Revolution EVO models. It aims to correlate in vivo TLD analysis with experimental CT Dose Index Volume (CTDIvol) values. The research reveals a maximum variation between machine-indicated CTDI values of 5.7% for the Optima 540 model and 6.8% for the Revolution EVO model at a voltage of 120 kV. Comparison with dose reference levels (DRL) from the United Kingdom and the United States indicates that the obtained values are below these standards, suggesting safe practices in the participating hospital. However, a larger sample size is recommended to establish local standards securely. In the thoracic region, nominal CTDI_vol values indicate measurements around 150% lower than the dose values measured by TLD's, however, no correlation was found between the two variables (p-value = 0.09). In the head, nominal CTDI_vol values varied on average 21% above the doses measured by TDL's, showing a strong correlation between the two quantities (p-value = 0.0002). The study highlights the importance of cautious interpretation of the CTDI_vol and the need for continuous optimization of procedures to ensure safe practices and minimize the risks of radiation exposure in pediatric patients.

Boron neutron capture therapy (BNCT) is a next-generation radiotherapy that combines neutron beams with boron compounds that selectively accumulate in cancer cells. Because this therapy requires neutrons for irradiation, compact accelerator-based neutron source devices that can be installed in hospitals are being developed around the world. Clinical trials of several devices have been conducted in Japan, China, and South Korea. The most advanced device was approved by the Japanese regulatory authorities in 2020. As a result, BNCT with the device is being implemented as an insured therapy for recurrent head and neck cancer at two hospitals in Japan. Thus, the development of accelerator-based neutron generators is proceeding in the field of BNCT. However, much remains to be done to establish this therapy as a common cancer treatment and to make it available worldwide. Dosimetry methods in BNCT are one of the issues. In the treatment, the dose given to a patient by neutron irradiation has to be estimated accurately. However, it is difficult to measure neutrons accurately and in real time. Currently, the neutron dose delivered to patients during treatment is not measured. Instead, the current of the charged particles that irradiate the neutron target material is measured to indirectly evaluate the neutron fluence and dose. In addition, accurate dose estimation based on reactions with neutrons and various elements in a human body uses the Monte Carlo method, which is computationally time-consuming. To address these dosimetry issues, several neutron fluence and dose measurement methods are being developed. This review briefly describes the current dosimetry methods and then presents several methods under development that allow real-time neutron measurements applicable to BNCT.

Current and emerging radiation therapy treatment techniques are characterized by complex and spatially non-uniform dose distributions with steep dose gradients delivered through high-precision dose placement with small radiation fields. Such approaches necessitate advanced verification tools. For accurate dose profile representation in stereotactic radiation therapy using megavoltage X-rays, the Nyquist theorem places the minimum sampling step as 2.5 mm that determines the required spatial resolution of the detectors for dose verification. For techniques such as micro- and mini-beam radiotherapy, micron-scale sampling is required, and sub-millimeter for particle therapy. Silicon semiconductor detectors provide the possibility of producing small sensitive volumes (including in arrays) without compromising mechanical robustness or sensitivity, therefore allowing for high spatial resolution dosimetry. They exhibit many properties of an ideal detector, while the properties of silicon detectors known to negatively influence response can be compensated for and minimized. As such, this work reviews the current status of semiconductor radiation detectors providing dose sampling by diodes with submillimeter sensitive volumes and a pitch of 2.5 mm or smaller for relative dosimetry in X-ray and electron external beam radiation therapy, brachytherapy, microbeam radiation therapy, proton, and heavy ion therapy. The following types of devices are discussed in this work, including commercial systems and those under development: one-dimensional and two-dimensional pixelated arrays (monolithic and single diode based).

This paper presents the evaluation of the fluence-to-ambient dose equivalent response function (referred to as F2D response function) of the Aloka TPS-451C neutron dose equivalent rate meter (hereafter referred to as a neutron meter) based on a singular value decomposition (SVD) approach. Measurements of neutron ambient dose equivalent rates, Ḣ${}^{\ast }\left(10\right)$, using the neutron meter were conducted in various neutron standard fields of ²⁴¹Am-Be source with different neutron fluence rate spectra. A series of Ḣ${}^{\ast }\left(10\right)$ values and corresponding neutron fluence rate spectra were used as the input data for unfolding the F2D response function of the neutron meter. To verify the SVD-based method, the fluence response functions of a well-known Bonner sphere spectrometer system were unfolded using the SVD method, and the results were compared with those provided in the IAEA technical report No. 403. The SVD method was then applied to determine the F2D response function of the Aloka TPS-451C neutron meter with the use of the ICRP-74 F2D response function and that predicted in a previous work as initial guesses. The consistency between initial guesses and the SVD unfolded F2D response function of the Aloka TPS-451C neutron meter implies the reliability of the SVD method for determining the response functions of neutron meters.

Biological dosimetry, using chromosome damage biomarkers, can be considered as a key measure for radiation overexposure assessment. Therefore, for accurate dose estimation through biodosimetry, it's imperative for each biological dosimetry laboratory to establish its own specific dose-response calibration curve. In this research, the cytokinesis-blocked micronucleus (CBMN) assay was utilized to determine the frequencies of micronuclei (MN) per binucleated cell in human peripheral blood lymphocytes exposed to x-ray radiation using a LINAC (6 MV) at doses up to 4 Gy. The aim was to establish an in vitro dose-response calibration curve in our laboratory for the Iranian demographic by analyzing blood samples of ten participants (5 males and 5 females) through CABAS and Dose Estimate software. Our findings indicate an over-dispersion of the Poisson distribution in the pooled data across both sexes, coupled with a linear-quadratic increase in MN yield with dose which was particularly pronounced in the female group. The constructed dose-response curves for micronuclei yield are represented by equations Y= (0.0102 ± 0.0016) +(0.0296 ± 0.0054) D+(0.0232 ± 0.0017) D² and Y= (0.0084 ± 0.0010) +(0.0212 ± 0.0034) D+(0.0160 ± 0.0011) D² for the female groups, respectively. The alpha and beta coefficients, derived from the Dose Estimate software for each male and female group, were closely comparable with those obtained from the CABAS program and previous studies. The analysis of statistical parameters such as the Weighted Chi Squared (χ2) and p-value suggests that using distinct curves for each sex, rather than a unified biodosimetry formula for pooled data, ensures more accurate radiation dose estimates. Consequently, the findings of this study provide us with the assurance to utilize the derived calibration curve of MN for upcoming biological dosimetry needs in Iran. However, to enhance the reliability of results, it's essential to use biodosimetry with diverse assays and consider all clinical and physical parameters, given the limitations of cytogenetic assays.

Animal models are essential in the development of new radiopharmaceuticals in nuclear medicine, particularly for accurate dose calculation in small animal internal dosimetry. This study presents a comprehensive dataset of S-values for eleven commonly used radionuclides, calculated using the DM_Bra mouse phantom with the GATE Monte Carlo simulation code. To validate our approach, we first compared S-values obtained from the DM_Bra phantom with published values derived from the Digimouse phantom using a Tc-99 m source. The differences between the two phantoms range from 0.68% to 12.45% for self-irradiation and from 0.15% to 4.19% for cross-irradiation when the source is the stomach. These results demonstrate good agreement with reference data, supporting the reliability of our dataset. We then expanded our analysis by generating S-values for additional radionuclides, reflecting their usage in both diagnostic and therapeutic applications. Furthermore, to assess the impact of varying mouse geometries on S-values, the DM_Bra phantom (26.9 g) was rescaled to represent two other mouse sizes (19.6 g and 35.9 g). The statistical uncertainty associated with all these S-values remains below 2%. This study offers a valuable resource for internal dosimetry in mice, providing detailed S-values for a wide range of radionuclides and organ geometries, which can be used in small animal PET and SPECT studies.

This study investigates in detail the effects of predose on TL and OSL signals for BeO dosimeters. The TL and OSL signals were deconvoluted to each peak and component. As a result of deconvolution, the variations in the kinetic parameters (E, T_max, b) for the TL signals and the lifetime values of the OSL signals were investigated. In addition, sensitivity changes according to predose were monitored for each peak and component. Finally, dose response curves were studied using the dose linearity index (f(D)) for each peak and component. Accordingly, the peak structures and kinetic parameters did not change according to the predose for the TL signal, whereas variations in the lifetime values for the OSL signal were observed, especially at the initial dose values (0.1 and 0.2 Gy). There was no change in sensitivity according to the predose for the total area condition although each peak and component exhibited independent behavior. Therefore, TL and OSL signals should be evaluated based on the total area in predose applications. The TL and OSL dose response curves exhibited different behaviors according to predose. TL dose response curves were not affected by the predose except for 1000 Gy, while the OSL dose response curves were affected by the predose considering the total area condition. The possible reason for the differences between the TL and OSL dose response curves is the significant transfer effect in the OSL signal at low doses, which results in greater changes at low doses compared to the TL signal. Also, thermal quenching effects may have resulted in lower intensity in the case of the TL signal. In future studies, preheating tests and thermal quenching corrections on TL peaks at high predoses may increase our understanding of deep trap interactions in BeO dosimeters.

This paper describes a method enabling the measurement of the potential alpha energy concentration (PAEC) of thoron decay products based on the determined concentration of lead ²¹²Pb (T_1/2 = 10.64 h) in the air [Lever et al., 2003]. A liquid scintillation spectrometer was used to determine the concentration of the ²¹²Pb isotope, and the sample was taken by pumping air through a filter where thoron decay products were stored. This method can be classified as integrating because the sample takes several hours, and the measurement results in one value for the entire sampling period. Measurements were carried out in laboratory conditions, in a climatic chamber where a constant supply of thoron was maintained, and in environmental conditions, in the basement of the family house and outdoors. Sampling took from 12 to 48 h. This article presents the preliminary results of the study. The obtained results were in the range of 170–195 Bq/m³ in the case of laboratory measurements and from 0.04 to 0.79 Bq/m³ in the case of environmental measurements. Based on the obtained results, the potential alpha energy concentration (PAEC) was calculated. The application of the low-level LS spectrometer allows for the achievement of a lower limit of detection (LLD) at level 0.04–0.05 Bq/m³, while the use of the portable LS spectrometer allows for the measurement of deficient ²¹²Pb concentrations in the range of 0.4–0.5 Bq/m³. The obtained results confirm that the method is suitable for determining the concentration of ²¹²Pb and, consequently, assessing exposure to thoron progeny.