Journal articles: 'Neutron dose'

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Hälg, Roger Antoine, and Uwe Schneider. "Neutron dose and its measurement in proton therapy—current State of Knowledge." British Journal of Radiology 93, no. 1107 (March 2020): 20190412. http://dx.doi.org/10.1259/bjr.20190412.

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Proton therapy has shown dosimetric advantages over conventional radiation therapy using photons. Although the integral dose for patients treated with proton therapy is low, concerns were raised about late effects like secondary cancer caused by dose depositions far away from the treated area. This is especially true for neutrons and therefore the stray dose contribution from neutrons in proton therapy is still being investigated. The higher biological effectiveness of neutrons compared to photons is the main cause of these concerns. The gold-standard in neutron dosimetry is measurements, but performing neutron measurements is challenging. Different approaches have been taken to overcome these difficulties, for instance with newly developed neutron detectors. Monte Carlo simulations is another common technique to assess the dose from secondary neutrons. Measurements and simulations are used to develop analytical models for fast neutron dose estimations. This article tries to summarize the developments in the different aspects of neutron dose in proton therapy since 2017. In general, low neutron doses have been reported, especially in active proton therapy. Although the published biological effectiveness of neutrons relative to photons regarding cancer induction is higher, it is unlikely that the neutron dose has a large impact on the second cancer risk of proton therapy patients.

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Fragopoulou, M., S. Stoulos, M. Manolopoulou, M. Krivopustov, and M. Zamani. "Dose Measurements around Spallation Neutron Sources." HNPS Proceedings 16 (January 1, 2020): 53. http://dx.doi.org/10.12681/hnps.2581.

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Neutron dose measurements and calculations around spallation sources are of importance for an appropriate shielding study. Two spallation sources, consisted of Pb target, have been irradiated by high-energy proton beams, delivered by the Nuclotron accelerator (JINR), Dubna. Dose measurements of the neutrons produced by the two spallation sources were performed using Solid State Nuclear Track Detectors (SSNTDs). In addition, the neutron dose after polyethylene and concrete was calculated using phenomenological model based on empirical relations applied in high energy Physics. Analytical and experimental neutron benchmark analysis has been performed using the transmission factor. A comparison of experimental results with calculations is given.

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Ivkovic, Ana, Dario Faj, Mladen Kasabasic, Marina Poje Sovilj, Ivana Krpan, Marina Grabar Branilovic, and Hrvoje Brkic. "The influence of shielding reinforcement in a vault with limited dimensions on the neutron dose equivalent in vicinity of medical electron linear accelerator." Radiology and Oncology 54, no. 2 (May 2, 2020): 247–52. http://dx.doi.org/10.2478/raon-2020-0024.

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AbstractBackgroundHigh energy electron linear accelerators (LINACs) producing photon beams with energies higher than 10 MeV are widely used in radiation therapy. In these beams, fast neutrons are generated, which results in undesired contamination of the therapeutic beam. In this study, measurements and Monte Carlo (MC) simulations were used to obtain neutron spectra and dose equivalents in vicinity of linear accelerator.Materials and methodsLINAC Siemens Oncor Expression in Osijek University Hospital is placed in vault that was previously used for 60Co machine. Then, the shielding of the vault was enhanced using lead and steel plates. Measurements of neutron dose equivalent around LINAC and the vault were done using CR-39 solid state nuclear track detectors. To compensate energy dependence of detectors, neutron energy spectra was calculated in measuring positions using MC simulations.ResultsThe vault is a source of photoneutrons, but a vast majority of neutrons originates from accelerator head. Neutron spectra obtained from MC simulations show significant changes between the measuring positions. Annual neutron dose equivalent per year was estimated to be less than 324 μSv in the measuring points outside of the vault.ConclusionsSince detectors used in this paper are very dependent on neutron energy, it is extremely important to know the neutron spectra in measuring points. Though, patient dosimetry should include neutrons, estimated annual neutron doses outside the vault were far below exposure limit of ionizing radiation for workers.

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Shagholi, Negin, Hassan Ali, Mahdi Sadeghi, Arjang Shahvar, Hoda Darestani, Banaee Nooshin, and Kheirolah Mohammadi. "Neutron dose evaluation of Elekta Linac at two energies (10 & 18 MV) by MCNP code and comparison with experimental measurements." JOURNAL OF ADVANCES IN PHYSICS 6, no. 1 (November 1, 2014): 1006–15. http://dx.doi.org/10.24297/jap.v6i1.1820.

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Medical linear accelerators, besides the clinically high energy electron and photon beams, produce other secondary particles such as neutrons which escalate the delivered dose. In this study the neutron dose at 10 and 18MV Elekta linac was obtained by using TLD600 and TLD700 as well as Monte Carlo simulation. For neutron dose assessment in 2020 cm2 field, TLDs were calibrated at first. Gamma calibration was performed with 10 and 18 MV linac and neutron calibration was done with 241Am-Be neutron source. For simulation, MCNPX code was used then calculated neutron dose equivalent was compared with measurement data. Neutron dose equivalent at 18 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 3.3, 4, 5 and 6 cm. Neutron dose at depths of less than 3.3cm was zero and maximized at the depth of 4 cm (44.39 mSvGy-1), whereas calculation resultedÂ in the maximum of 2.32 mSvGy-1 at the same depth. Neutron dose at 10 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 2.5, 3.3, 4 and 5 cm. No photoneutron dose was observed at depths of less than 3.3cm and the maximum was at 4cm equal to 5.44mSvGy-1, however, the calculated data showed the maximum of 0.077mSvGy-1 at the same depth. The comparison between measured photo neutron dose and calculated data along the beam axis in different depths, shows that the measurement data were much more than the calculated data, so it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry in linac central axis due to high photon flux, whereas MCNPX Monte Carlo techniques still remain a valuable tool for photonuclear dose studies.

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Gutermuth, F., T. Radon, G. Fehrenbacher, and J. G. Festag. "The response of various neutron dose meters considering the application at a high energy particle accelerator." Kerntechnik 68, no. 4 (August 1, 2003): 172–79. http://dx.doi.org/10.1515/kern-2003-0072.

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Abstract The applicability of several neutron detectors for dose measurements at a neutron field typical for high energy particle accelerators is investigated. The response of four commercially available active neutron dose meters and two passive detectors to neutrons from a 241Am-Be(α,n) source and to neutrons at the CERN EU high energy reference field was determined experimentally and simulated using the Monte-Carlo code FLUKA. Fluence response functions and dose responses for the different detectors were calculated in the energy range between 1 keV and 10 GeV. The results show that the dose response to the high energy neutron field at CERN of the conventional rem-counters is lower by a factor of 2 to 2.5 if compared to the dose response to a 241Am-Be(α,n) neutron source. The rem-counters exhibiting an additional layer of lead inside the moderating structure showed dose readings which differ only up to 25 %. A thermoluminescent based neutron detector was tested for comparison. These passive detectors revealed a neutron response similar to a rem-counter and may be preferable for situations with highly pulsed beams.

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Dennis, J. A., and L. A. Dennis. "Neutron dose effect relationships at low doses." Radiation and Environmental Biophysics 27, no. 2 (June 1988): 91–101. http://dx.doi.org/10.1007/bf01214599.

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Zhou, Bin, Fei Shen, Zhiliang Hu, Songlin Wang, Xichao Ruan, and Tianjiao Liang. "A Study of Stray Neutron Field Measurements for the Neutron Scattering Instruments at CSNS." Applied Sciences 12, no. 10 (May 12, 2022): 4915. http://dx.doi.org/10.3390/app12104915.

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Stray neutrons might cause several negative impacts. However, it is usually difficult to conduct precise stray neutron simulations using the Monte Carlo method. Therefore, in this study, a measurement technique was proposed to study the stray neutrons experimentally inside the neutron scattering instruments at China Spallation Neutron Source (CSNS). The adopted measurement instruments comprise an extended-range Bonner sphere spectrometer and a commercial neutron ambient-dose-equivalent dosimeter, which enables us to directly measure the neutron spectra and ambient-dose equivalent H*(10) values. Verification experiments were performed inside the BL06 beam line experimental area at CSNS at two exposed locations with different sample conditions. Comparison of the experimentally measured neutron spectra, integral neutron fluence, and H*(10) value with the simulations demonstrated the feasibility of using the proposed method for studying stray neutrons for the neutron instruments.

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D’Avino, Vittoria, Fabrizio Ambrosino, Roberto Bedogni, Abner Ivan C. Campoy, Giuseppe La Verde, Silvia Vernetto, Carlo Francesco Vigorito, and Mariagabriella Pugliese. "Characterization of Thermoluminescent Dosimeters for Neutron Dosimetry at High Altitudes." Sensors 22, no. 15 (July 30, 2022): 5721. http://dx.doi.org/10.3390/s22155721.

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Neutrons constitute a significant component of the secondary cosmic rays and are one of the most important contributors to natural cosmic ray radiation background dose. The study of the cosmic ray neutrons’ contribution to the dose equivalent received by humans is an interesting and challenging task for the scientific community. In addition, international regulations demand assessing the biological risk due to radiation exposure for both workers and the general population. Because the dose rate due to cosmic radiation increases significantly with altitude, the objective of this work was to characterize the thermoluminescent dosimeter (TLDs) from the perspective of exposing them at high altitudes for longtime neutron dose monitoring. The pair of TLD-700 and TLD-600 is amply used to obtain the information on gamma and neutron dose in mixed neutron-gamma fields due to the present difference in 6Li isotope concentration. A thermoluminescence dosimeter system based on pair of TLD-600/700 was characterized to enable it for neutron dosimetry in the thermal energy range. The system was calibrated in terms of neutron ambient dose equivalent in an experimental setup using a 241Am-B radionuclide neutron source coated by a moderator material, polyethylene, creating a thermalized neutron field. Afterward, the pair of TLD-600/700 was exposed at the CERN-EU High-Energy Reference Field (CERF) facility in Geneva, which delivers a neutron field with a spectrum similar to that of secondary cosmic rays. The dosimetric system provided a dose value comparable with the calculated one demonstrating a good performance for neutron dosimetry.

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Konstantin Andreevich Kuznetsov, Pavel Semenovich Kizim, Andrey Yurievich Berezhnoy, Oleksandr Pilipovich Shchus, and Gennadiy Michailovich Onyshchenko. "Cell stress response to low-dose neutron radiation." Magna Scientia Advanced Biology and Pharmacy 1, no. 1 (November 30, 2020): 036–42. http://dx.doi.org/10.30574/msabp.2020.1.1.0022.

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Background. It is a point of discussion whether low-dose ionizing radiation has harmful or stimulating impact on cell. According to high relative biological effectiveness of neutron radiation there is a need of description of any process triggered in the cell by neutrons. Objective. The aim of current work is the investigation of the low dosed neutron radiation effects on human cells by indicators of cell stress such as state of chromatin and cell membrane permeability. Materials and methods. Human buccal epithelium cells from 3 male donors (21, 24, 25 years old) were exposed to fast neutron radiation in dose range 2.3–146.0 mSv from 239Pu-Be source. State of chromatin was evaluated by count of heterochromatin granules quantity in 100 nuclei stained with 2% orcein in 45% acetic acid; ratio of cells with increased membrane permeability stained with 5 mM indigocarmine in 300 cells. Results. Changes in level of heterochromatin granules quantity and in cell membrane permeability revealed wave-shaped dependency with maximum effects at 36.5 mSv. Further increase of dose resulted in return of both chromatin state and membrane permeability levels closely to control or even lower. Conclusion. Membrane restoration and chromatin decompaction under doses higher than 36.5 mSv together can be a sign of hormetic (stimulating) effect of low-dose neutron radiation.

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Chen, Zhao, Peng Yang, Qin Lei, Yumei Wen, Donglin He, Zhangwen Wu, and Chengjun Gou. "COMPARISON OF BNCT DOSIMETRY CALCULATIONS USING DIFFERENT GEANT4 PHYSICS LISTS." Radiation Protection Dosimetry 187, no. 1 (May 28, 2019): 88–97. http://dx.doi.org/10.1093/rpd/ncz144.

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Abstract A comparison of Geant4 physics lists is conducted in the calculation of the total absorbed dose, boron dose, and non-boron dose in phantom, and the total depth-dose, boron depth-dose, and non-boron depth-dose along the beam axis for neutrons in a range of 0.0253 eV to 10 MeV. Physics processes are included for neutrons, photons, and charged particles, and calculations are conducted for neutrons and secondary particles. The results obtained from QBBC, QGSP_BERT, and neutron high precision physics lists with and without S(α, β) data are compared with the FLUKA values. Neutron high precision physics lists with S(α, β) data showed the best agreement with FLUKA in the studied energy range.

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Matsumoto, Shinnosuke, and Shunsuke Yonai. "EVALUATION OF NEUTRON AMBIENT DOSE EQUIVALENT IN INTENSITY-MODULATED COMPOSITE PARTICLE THERAPY." Radiation Protection Dosimetry 193, no. 2 (January 2021): 90–95. http://dx.doi.org/10.1093/rpd/ncab031.

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Abstract Several studies have reported benefits derived from cancer treatment using various heavy-ion beams. Based on these reports, the National Institutes for Quantum and Radiological Science and Technology started developing intensity-modulated composite particle therapy (IMPACT) using He-, C-, O-, and Ne-ions. In ion beam therapy, nuclear interactions in the beamline devices or patient produce secondary neutrons. This study evaluated the characteristics of secondary neutrons in IMPACT. Neutron ambient dose equivalents were measured using WENDI-II. Measurements were performed under realistic case scenarios using He-, C-, O- and Ne-ion beams. Moreover, neutron ambient dose equivalents generated by He-, C-, O- and Ne-ion beams were compared with neutron ambient dose equivalents in proton therapy. No differences exist in the distance-dependence even when the primary ions are different. Neutrons generated by primary ion beams of high atomic numbers tend to emit forward. Moreover, in contrast with proton therapy, IMPACT can reduce neutron doses.

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Fedorov, S. G., A. V. Berlyand, and V. M. Dyachenko. "Ensuring the unity of measurements of the neutron radiation rate absorbed dose in the field of clinical neutron radiation dosimetry." Journal of Physics: Conference Series 2373, no. 2 (December 1, 2022): 022046. http://dx.doi.org/10.1088/1742-6596/2373/2/022046.

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Abstract Development of neutron radiation therapy arises the question of ensuring the uniformity of measurements of the absorbed dose and the absorbed dose rate of neutron radiation. FGUP “VNIIFTRI” improved the primary method and means of reproducing the unit of neutron radiation absorbed dose rate within the framework of improving the State primary standard of units of absorbed dose rate and neutron dose equivalent rate GET 117-2010. A set of ionization chambers has been created and the upper value of the reproduction of the unit of absorbed dose rate of neutron radiation has been expanded. A hardware-methodical complex has also been developed for transferring a unit of absorbed dose rate of neutron radiation to nuclear physics facilities used in neutron radiation therapy. As a result the expanded relative uncertainty of reproduction of the unit of absorbed dose rate does not exceed 3%, the expanded relative uncertainty of the transmission method does not exceed 0.5%. The resulting level of accuracy is in line with the recommendations used in neutron beam therapy.

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Suzuki, Chihiro, Takuya Abe, Takeshi Iimoto, Ken-ichi Tanaka, Hidenori Tanabe, and Toshiso Kosako. "ICONE19-44007 EFFECT OF SURROUNDING ENVIRONMENT TO NEUTRON SKYSHINE DOSE FROM A FAST NEUTRON SOURCE REACTOR." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2011.19 (2011): _ICONE1944. http://dx.doi.org/10.1299/jsmeicone.2011.19._icone1944_4.

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Arianto, Fajar, Liska Tri Handayani, Wahyu Setia Budi, and Prasetyo Basuki. "Determination of Neutron Flux in Brain Cancer Boron Neutron Capture Therapy Using Monte Carlo Simulation." Physics Communication 6, no. 2 (November 30, 2022): 79–84. http://dx.doi.org/10.15294/physcomm.v6i2.40277.

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Boron Neutron Capture Therapy (BNCT) is a relatively safer technology for killing cancer cells, one of which is the Glioblastoma multiforme. One of the main components of the BNCT equipment is the collimator which functions as an exit point for epithermal neutron particles that hit cancer cells. In addition to the experimental method, BNCT research can be carried out by modeling, including using the MCNPX software based on the Monte Carlo Method. This research aimed to determine the flux distribution of fast and epithermal neutrons and the dose rate of fast neutrons and gamma that hit the target cancer cells in the phantom head of ORNL MIRD. Modeling using the MCNPX software has three main parts: cell cards, surface cards, and data cards. A tally is used on the data card to calculate the neutron flux. Based on the calculation of the modeling results, the flux of epithermal neutron is 2.87 x 109 n/cm2.s. The dose ratio of the epithermal to the fast neutron flux is 2.29 x 10-14 Gy.cm2/n. Then, the balance of the dose rate of the epithermal to the gamma is 1.64 x 10-14 Gy.cm2/n, and the ratio of epithermal to thermal neutron flux is 0.004. In this study, the epithermal neutron flux hitting the target cancer cells in cell target was moderated at 4 cm so that at a depth of 8 cm, the energy was converted into thermal neutrons. Based on the analysis of the results, it can be concluded that the neutron flux that will interact with cancer tissue is thermal neutrons, not epithermal neutron flux.

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Savitri, Leily, Iswandarini Iswandarini, and Rusmanto Rusmanto. "Kajian pemantauan dosis neutron terhadap pekerja radiasi pada pengoperasian Linac dengan energi foton ≥ 10 MV." Jurnal Pengawasan Tenaga Nuklir 1, no. 2 (December 15, 2021): 47–55. http://dx.doi.org/10.53862/jupeten.v1i2.018.

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The use of Linac for radiotherapy is starting to use a lot of high-energy photons of 10 MV; in addition, some use 15 MV for patient therapy in routine use, there is also the use of 6 MV. The purpose of this study is to obtain an overview and information of the neutron dose that has the potential to provide additional doses for radiation workers operating the Linac 10 MV aircraft. Based on the Regulation of the Head of BAPETEN No. 3 of 2013, Article 48 paragraph (2) states that in the operation of Linac with X-ray photon energies above 10 MV, must coat the shield wall with a neutron-absorbing material. The statement follows the IAEA-TecDoc 1891 that neutrons will have the potential to have a significant radiological impact on workers if routinely operated at energies above 10 MV, so must consider protection for workers. The results of a survey from 27 hospitals, obtained information through filling out questionnaires and discussions and validated with B@LIS Pendora, it found that the trend of annual doses received by each profession in the operation of Linac 6 MV, 10 MV, and 15 MV was less than one mSv, only partially small worker dose that is above one mSv (above the 90th percentile). This study concluded that the presence of neutrons in Linacs up to 10 MV was deemed not to have a significant radiological impact on workers. The recommended criteria/mechanism for monitoring worker neutron doses in Linacs up to 10 MV could be based on if the safety study results obtained a dose received by workers 1.5 mSv/year. Then, there is no need to monitor the neutron dose. If the measurement results of exposure to neutron and gama radiation around the Linac space are 7.5 microSv/hour, there is no need for neutron monitoring. In Linacs above 10 MV, if the annual effective dose is 1.5 mSv/year, there is no need to monitor the dose of special neutron personnel. Still, routine radiation exposure monitoring may be considered every two years. Keywords: Neutron Dose, Radiation Worker, Linac, Dose Monitoring.

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Krstic, Dragana, Dragoslav Nikezic, Milena Zivkovic, and Marija Jeremic. "Dose assessment with MCNP5/X code for boron neutron capture therapy of pancreas cancer." Nuclear Technology and Radiation Protection 36, no. 3 (2021): 294–98. http://dx.doi.org/10.2298/ntrp2103294k.

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Boron neutron capture therapy is based on neutron capture by 10B and creation of high energy alpha particle and recoiled 7Li ion which have ranges comparable to cell dimensions. The MCNP5/X code is applied for calculation of the absorbed doses as well as dose distribution by depth in pancreas cancer and the organs of the analytical and voxelized Oak Ridge National Laboratory phantom, which represents the human body. The depth dose distributions for thermal, epithermal neutrons, and neutron spectrum, show that the absorbed doses are the largest exactly in the cancer, in all cases. It is found by this simulation that the epithermal neutrons are the optimal choice for BNC therapy. They deliver a similar dose to cancer as the thermal neutrons but, spare the healthy tissue more than the thermal ones.

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Choynzonov, E. L., V. A. Lisin, Zh A. Gribova, V. V. Velikaya, and O. V. Startseva. "Methodological approaches to prevention of radiation-induced skin reactions in neutron-photon therapy for malignant neoplasms." Siberian journal of oncology 18, no. 2 (April 26, 2019): 44–51. http://dx.doi.org/10.21294/1814-4861-2019-18-2-44-51.

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The goal of radiotherapy is to maximize the radiation dose to abnormal cancer cells while preventing damage to healthy tissue. in neutron therapy, the optimum regime of treatment is uncertain to date.The purpose of the study to develop a set of methodological approaches that ensure the permissible frequency and severity of radiation-induced reactions in cancer patients subjected to neutron and neutron-photon therapy (NFt) using u-120 cyclotron.Material and methods. We used the dependence of the relative biological effectiveness (RBE) of neutrons on the dose and time-dose-fractionation model (tdF). the interaction of neutrons with various types of tissues was analyzed, and the algorithm for summing neutron and photon doses in neutronphoton therapy was developed.Results. Clinical studies of neutron-photon therapy showed that the developed approaches can predict and prevent serious damage to normal tissue with a satisfactory accuracy. the role of all factors influencing the nature of radiation reactions was taken into account in the computer program, which allowed the main characteristics of the planned courses of neutron-photon therapy to be obtained.

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Mohammadi, Sara, Marziyeh Behmadi, Aghil Mohammadi, and Mohammad Taghi Bahreyni Toossi. "THERMAL AND FAST NEUTRON DOSE EQUIVALENT DISTRIBUTION MEASUREMENT OF 15-MV LINEAR ACCELERATOR USING A CR-39 NUCLEAR TRACK DETECTORS." Radiation Protection Dosimetry 188, no. 4 (February 5, 2020): 503–7. http://dx.doi.org/10.1093/rpd/ncaa001.

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Abstract The main purpose of this study is to measure the contribution of the thermal and fast neutron dose along the central axis of the 15 MV Elekta Precise linac in a tissue equivalent phantom. In order to achieve this purpose, different points were selected in three field sizes of 5 × 5 cm2, 10 × 10 cm2 and 15 × 15 cm2. Fast and thermal neutrons were measured using CR-39 nuclear track detectors with and without thermal neutron converter of 10B, respectively. According to the results, the fast neutron dose equivalent was decreased as the depth increased (field size 5 × 5, 10 × 10 and 15 × 15 cm2 fall from 0.35 to 0.15, 0.5 to 0.3 and 0.5 to 0.3, respectively). Thermal dose equivalent was increased as the depth increased in the tissue equivalent phantom (field size 5 × 5, 10 × 10 and 15 × 15 cm2 rise from 0.1 to 0.4, 0.4 to 0.8 and 0.4 to 0.9, respectively). In conclusion, at depth <3 cm, most existing neutrons are fast and CR-39 films are sensitive to fast neutrons; therefore, they are more appropriate than thermoluminescent dosemeters in measuring neutron dose equivalent.

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Maysaroh, Atika, Kusminarto Kusminarto, Dwi Satya Palupi, and Yohannes Sardjono. "Dosimetry of In Vivo Experiment for Lung Cancer Based on Boron Neutron Capture Therapy on Radial Piercing Beam Port Kartini Nuclear Reactor by MCNPX Simulation Method." ASEAN Journal on Science and Technology for Development 35, no. 3 (December 24, 2018): 213–16. http://dx.doi.org/10.29037/ajstd.540.

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Cancer is one of the leading causes of death globally, with lung cancer being among the most prevalent. Boron Neutron Capture Therapy (BNCT) is a cancer therapy method that uses the interaction between thermal neutrons and boron-10 which produces a decaying boron-11 particle and emits alpha, lithium 7 and gamma particles. A study was carried out to model an in vivo experiment of rat organisms that have lung cancer. Dimensions of a rat’s body were used in Konijnenberg research. Modeling lung cancer type, non-small cell lung cancer, was used in Monte Carlo N Particle-X. Lung cancer was modeled with a spherical geometry consisting of 3 dimensions: PTV, GTV, and CTV. In this case, the neutron source was from the radial piercing beam port of Kartini Reactor, Yogyakarta. The variation of boron concentration was 20, 25, 30, 35, 40, and 40 µg/g cancer. The output of the MCNP calculation was neutron scattering dose, gamma-ray dose and neutron flux from the reactor. A neutron flux was used to calculate the alpha proton and gamma-ray dose from the interaction of tissue material and thermal neutrons. The total dose was calculated from a four-dose component in BNCT. The results showed that the dose rate will increase when the boron concentration is higher, whereas irradiating time will decrease.

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Sovilj, Marina Poje, Branko Vuković, Vanja Radolić, Igor Miklavčić, and Denis Stanić. "Potential benefit of retrospective use of neutron monitors in improving ionising radiation exposure assessment on international flights: issues raised by neutron passive dosimeter measurements and EPCARD simulations during sudden changes in solar activity." Archives of Industrial Hygiene and Toxicology 71, no. 2 (June 1, 2020): 152–57. http://dx.doi.org/10.2478/aiht-2020-71-3403.

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AbstractSince air transport became more accessible, more and more people have been exposed to ionising radiation of cosmic origin. Measuring the neutron dose equivalent is a good approximation of total ambient dose equivalent, as neutrons carry about 50 % of the dose at flight altitudes. The aim of our study was to compare our measurements of the neutron component of secondary cosmic radiation dose, taken with passive dosimeters, with the data obtained from a simulation generated by EPCARD software, which is common in assessing flight crew exposure to ionising radiation. We observed deviations (both above and below) from the expected proportion of the neutron component (between 40 and 80 %), which pointed to certain issues with actual passive dosimeter measurement and the EPCARD simulation. The main limitation of the dosimeter are large uncertainties in high energy neutron response, which may result in underestimation of neutron dose equivalent. The main drawback of the software simulation is monthly averaging of solar potential in calculations, which can neglect sporadic high energy events. Since airlines worldwide almost exclusively use software (due to costs and convenience) to estimate the dose received by their crew, it is advisable to retrospectively recalculate the dose taking into account neutron monitor readings when solar activity changes.

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Penner, Crystal, Samuel Usherovich, Jana Niedermeier, Camille Belanger-Champagne, Michael Trinczek, Elisabeth Paulssen, and Cornelia Hoehr. "Organic Scintillator-Fibre Sensors for Proton Therapy Dosimetry: SCSF-3HF and EJ-260." Electronics 12, no. 1 (December 20, 2022): 11. http://dx.doi.org/10.3390/electronics12010011.

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In proton therapy, the dose from secondary neutrons to the patient can contribute to side effects and the creation of secondary cancer. A simple and fast detection system to distinguish between dose from protons and neutrons both in pretreatment verification as well as potentially in vivo monitoring is needed to minimize dose from secondary neutrons. Two 3 mm long, 1 mm diameter organic scintillators were tested for candidacy to be used in a proton–neutron discrimination detector. The SCSF-3HF (1500) scintillating fibre (Kuraray) and EJ-260 plastic scintillator (Eljen Technology) were irradiated at the TRIUMF Neutron Facility and the Proton Therapy Research Centre. In the proton beam, we compared the raw Bragg peak and spread-out Bragg peak response to the industry standard Markus chamber detector. Both scintillator sensors exhibited quenching at high LET in the Bragg peak, presenting a peak-to-entrance ratio of 2.59 for the EJ-260 and 2.63 for the SCSF-3HF fibre, compared to 3.70 for the Markus chamber. The SCSF-3HF sensor demonstrated 1.3 times the sensitivity to protons and 3 times the sensitivity to neutrons as compared to the EJ-260 sensor. Combined with our equations relating neutron and proton contributions to dose during proton irradiations, and the application of Birks’ quenching correction, these fibres provide valid candidates for inexpensive and replicable proton-neutron discrimination detectors.

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Ramadhani, Amanda Dhyan Purna, Susilo Susilo, Irfan Nurfatthan, Yohannes Sardjono, Widarto Widarto, Gede Sutresna Wijaya, and Isman Mulyadi Triatmoko. "DOSE ESTIMATION OF THE BNCT WATER PHANTOM BASED ON MCNPX COMPUTER CODE SIMULATION." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 22, no. 1 (March 25, 2020): 23. http://dx.doi.org/10.17146/tdm.2020.22.1.5780.

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Cancer is a malignant tumor that destroys healthy cells. Cancer treatment can be done by several methods, one of which is BNCT. BNCT uses 10B target which is injected into the human body, then it is irradiated with thermal or epithermal neutrons. Nuclear reaction will occur between boron and neutrons, producing alpha particle and lithium-7. The dose is estimated by how much boron and neutron should be given to the patient as a sum of number of boron, number of neutrons, number of protons, and number of gamma in the reaction of the boron and neutron. To calculate the dose, the authors simulated the reaction with Monte Carlo N Particle-X computer code. A water phantom was used to represent the human torso, as 75% of human body consists of water. Geometry designed in MCNPX is in cubic form containing water and a cancer cell with a radius of 2 cm. Neutron irradiation is simulated as originated from Kartini research reactor, modeled in cylindrical form to represent its aperture. The resulting total dose rate needed to destroy the cancer cell in GTV is 2.0814×1014 Gy.s (76,38%) with an irradiation time of 1,4414×10-13 s. In PTV the dose is 5.2295×1013 Gy.s (19,19%) with irradiation time of 5.7367×10-13 s. In CTV, required dose is 1.1866×1013 Gy.s (4,35%) with an irradiation time of 2.5283×10-12 s. In the water it is 1.9128×1011 Gy.s (0,07%) with an irradiation time of 1,5684×10-10 s. The irradiation time is extremely short since the modeling is based on water phantom instead of human body.Keywords: BNCT, Dose, Cancer, Water Phantom, MCNPX

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Pyshkina, Mariya, Mihail Zhukovskiy, Aleksey Vasil'ev, and Marina Romanova. "Oral Thermoluminescent Neutron Dosimeter for Emergency Exposure Conditions." ANRI, no. 2 (June 29, 2021): 65–74. http://dx.doi.org/10.37414/2075-1338-2021-105-2-65-74.

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An oral dosimeter of mixed gamma-neutron radiation for emergency exposure conditions has been developed. The energy dependence of the neutron radiation dosimeter sensitivity is close to the energy dependence of the specific effective dose per unit flux density. For neutron fields containing a significant contribution of fast neutrons, the uncertainty of the dosimeter readings is no more than 25% for the anteroposterior radiation geometry and no more than 35% for the rotation geometry. In neutron fields with a predominance of particles with thermal and intermediate energies, the dosimeter overestimates the effective radiation dose by 2.5 times for the anteroposterior geometry and 3.3 times for the rotation geometry. A staging experiment was carried out, which included placing individual dosimeters inside a canister simulating the torso of a standard adult in a neutron radiation field. The conditionally true values of the effective dose were obtained using the energy and angular distribution of the neutron radiation flux density. Differences in the dosimeter readings and the conditionally true value of the effective dose do not exceed 2.

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Ng, Candy Y. P., Eva Y. Kong, Alisa Kobayashi, Noriyoshi Suya, Yukio Uchihori, Shuk Han Cheng, Teruaki Konishi, and Kwan Ngok Yu. "Non-induction of radioadaptive response in zebrafish embryos by neutrons." Journal of Radiation Research 57, no. 3 (June 1, 2016): 210–19. http://dx.doi.org/10.1093/jrr/rrv089.

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Abstract In vivo neutron-induced radioadaptive response (RAR) was studied using zebrafish ( Danio rerio ) embryos. The Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility at the National Institute of Radiological Sciences (NIRS), Japan, was employed to provide 2-MeV neutrons. Neutron doses of 0.6, 1, 25, 50 and 100 mGy were chosen as priming doses. An X-ray dose of 2 Gy was chosen as the challenging dose. Zebrafish embryos were dechorionated at 4 h post fertilization (hpf), irradiated with a chosen neutron dose at 5 hpf and the X-ray dose at 10 hpf. The responses of embryos were assessed at 25 hpf through the number of apoptotic signals. None of the neutron doses studied could induce RAR. Non-induction of RAR in embryos having received 0.6- and 1-mGy neutron doses was attributed to neutron-induced hormesis, which maintained the number of damaged cells at below the threshold for RAR induction. On the other hand, non-induction of RAR in embryos having received 25-, 50- and 100-mGy neutron doses was explained by gamma-ray hormesis, which mitigated neutron-induced damages through triggering high-fidelity DNA repair and removal of aberrant cells through apoptosis. Separate experimental results were obtained to verify that high-energy photons could disable RAR. Specifically, 5- or 10-mGy X-rays disabled the RAR induced by a priming dose of 0.88 mGy of alpha particles delivered to 5-hpf zebrafish embryos against a challenging dose of 2 Gy of X-rays delivered to the embryos at 10 hpf.

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Niedermeier, Jana, Crystal Penner, Samuel Usherovich, Camille Bélanger-Champagne, Elisabeth Paulssen, and Cornelia Hoehr. "Optical Fibers as Dosimeter Detectors for Mixed Proton/Neutron Fields—A Biological Dosimeter." Electronics 12, no. 2 (January 8, 2023): 324. http://dx.doi.org/10.3390/electronics12020324.

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In recent years, proton therapy has gained importance as a cancer treatment modality due to its conformality with the tumor and the sparing of healthy tissue. However, in the interaction of the protons with the beam line elements and patient tissues, potentially harmful secondary neutrons are always generated. To ensure that this neutron dose is as low as possible, treatment plans could be created to also account for and minimize the neutron dose. To monitor such a treatment plan, a compact, easy to use, and inexpensive dosimeter must be developed that not only measures the physical dose, but which can also distinguish between proton and neutron contributions. To that end, plastic optical fibers with scintillation materials (Gd2O2S:Tb, Gd2O2S:Eu, and YVO4:Eu) were irradiated with protons and neutrons. It was confirmed that sensors with different scintillation materials have different sensitivities to protons and neutrons. A combination of these three scintillators can be used to build a detector array to create a biological dosimeter.

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Chang, Insu, Sang In Kim, Jung Il Lee, Jang Lyurl Kim, and Bong Hwan Kim. "Neutron Dose Measurements Using TLDs in a252Cf Neutron Field." Journal of Radiation Protection 38, no. 1 (March 30, 2013): 37–43. http://dx.doi.org/10.14407/jrp.2013.38.1.037.

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Mousavi Shirazi, Seyed Alireza, and Dariush Sardari. "Design and Simulation of a New Model for Treatment by NCT." Science and Technology of Nuclear Installations 2012 (2012): 1–7. http://dx.doi.org/10.1155/2012/213640.

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In this investigation, neutron capture therapy (NCT) through high energy neutrons using Monte Carlo method has been studied. In this study a new method of NCT for a sample liver phantom has been defined, and interaction of 12 MeV neutrons with a multilayer spherical phantom is considered. In order to reach the desirable energy range of neutrons in accord with required energy in absence of eligible clinical neutron source for NCT, this model of phantom might be utilized. The neutron flux and the deposited dose in the all components and different layers of the mentioned phantom are computed by Monte Carlo simulation. The results of Monte Carlo method are compared with analytical method results so that by using a computer program in Turbo-Pascal programming, the deposited dose in the liver phantom has been computed.

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Nair, Shankari, Monique Engelbrecht, Xanthene Miles, Roya Ndimba, Randall Fisher, Peter du Plessis, Julie Bolcaen, Jaime Nieto-Camero, Evan de Kock, and Charlot Vandevoorde. "The Impact of Dose Rate on DNA Double-Strand Break Formation and Repair in Human Lymphocytes Exposed to Fast Neutron Irradiation." International Journal of Molecular Sciences 20, no. 21 (October 28, 2019): 5350. http://dx.doi.org/10.3390/ijms20215350.

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The lack of information on how biological systems respond to low-dose and low dose-rate exposures makes it difficult to accurately assess the carcinogenic risks. This is of critical importance to space radiation, which remains a serious concern for long-term manned space exploration. In this study, the γ-H2AX foci assay was used to follow DNA double-strand break (DSB) induction and repair following exposure to neutron irradiation, which is produced as secondary radiation in the space environment. Human lymphocytes were exposed to high dose-rate (HDR: 0.400 Gy/min) and low dose-rate (LDR: 0.015 Gy/min) p(66)/Be(40) neutrons. DNA DSB induction was investigated 30 min post exposure to neutron doses ranging from 0.125 to 2 Gy. Repair kinetics was studied at different time points after a 1 Gy neutron dose. Our results indicated that γ-H2AX foci formation was 40% higher at HDR exposure compared to LDR exposure. The maximum γ-H2AX foci levels decreased gradually to 1.65 ± 0.64 foci/cell (LDR) and 1.29 ± 0.45 (HDR) at 24 h postirradiation, remaining significantly higher than background levels. This illustrates a significant effect of dose rate on neutron-induced DNA damage. While no significant difference was observed in residual DNA damage after 24 h, the DSB repair half-life of LDR exposure was slower than that of HDR exposure. The results give a first indication that the dose rate should be taken into account for cancer risk estimations related to neutrons.

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Obeid, M. H., A. Ismail, A. Bitar, and R. Shweikani. "ESTIMATION OF SCATTERED NEUTRONS CONTRIBUTION IN A NEUTRON CALIBRATION BUNKER USING A MONTE CARLO SIMULATION." Radiation Protection Dosimetry 198, no. 1-2 (January 2022): 37–43. http://dx.doi.org/10.1093/rpd/ncab184.

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Abstract The contribution of scattered neutrons is inevitable in neutron calibration facilities. This contribution complicates the measurements of neutron radiation, therefore, it should be estimated to correct the response of neutron probe instruments. In the present work, Monte Carlo simulation was performed for a neutron calibration bunker using the MCNP-4C code. This simulation aimed to calculate the contribution ratio of scattered neutrons to the neutron field. To simulate the neutron field, 241Am-Be neutron source defined in the ISO 8529-1 was used. The results of the simulation reported in this work were found to be consistent with those found experimentally in previous work. Additionally, the distribution of both the ambient dose equivalent rate and the contribution ratio of scattered neutrons in the bunker were mapped using this simulation.

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Elmtalab, Soheil, Iraj Abedi, Zahra Alirezaei, Mohammad Hossein Choopan Dastjerdi, Ghazale Geraily, and Amir Hossein Karimi. "Semi-experimental assessment of neutron equivalent dose and secondary cancer risk for off-field organs in glioma patients undergoing 18-MV radiotherapy." PLOS ONE 17, no. 7 (July 29, 2022): e0271028. http://dx.doi.org/10.1371/journal.pone.0271028.

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Neutron contamination as a source of out-of-field dose in radiotherapy is still of concern. High-energy treatment photons have the potential to overcome the binding energy of neutrons inside the nuclei. Fast neutrons emitting from the accelerator head can directly reach the patient’s bed. Considering that modern radiotherapy techniques can increase patient survival, concerns about unwanted doses and the lifetime risk of fatal cancer remain strong or even more prominent, especially in young adult patients. The current study addressed these concerns by quantifying the dose and risk of fatal cancer due to photo-neutrons for glioma patients undergoing 18-MV radiotherapy. In this study, an NRD model rem-meter detector was used to measure neutron ambient dose equivalent, H*(10), at the patient table. Then, the neutron equivalent dose received by each organ was estimated concerning the depth of each organ and by applying depth dose corrections to the measured H*(10). Finally, the effective dose and risk of secondary cancer were determined using NCRP 116 coefficients. Evidence revealed that among all organs, the breast (0.62 mSv/Gy) and gonads (0.58 mSv/Gy) are at risk of photoneutrons more than the other organs in such treatments. The neutron effective dose in the 18-MV conventional radiotherapy of the brain was 13.36 mSv. Among all organs, gonads (6.96 mSv), thyroid (1.86 mSv), and breasts (1.86 mSv) had more contribution to the effective dose, respectively. The total secondary cancer risk was estimated as 281.4 cases (per 1 million persons). The highest risk was related to the breast and gonads with 74.4 and, 34.8 cases per 1 million persons, respectively. Therefore, it is recommended that to prevent late complications (secondary cancer and genetic effects), these organs should be shielded from photoneutrons. This procedure not only improves the quality of the patient’s personal life but also the healthy childbearing in the community.

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Musabaeva, Ludmila I., and Valery A. Lisin. "Response of Resistant Malignant Tumors to Neutron Therapy." Advanced Materials Research 1084 (January 2015): 467–70. http://dx.doi.org/10.4028/www.scientific.net/amr.1084.467.

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Tumor response to 6.3 MeV fast neutrons generated by U-120 cyclotron was assessed. The study included: 26 patients with locally advanced breast cancer, who received neutron therapy in the regimen of high-dose daily fractions, followed by surgery; 41patients with head and neck cancer and 44 patients with musculoskeletal system cancer. The devised method of 6.3 MeV neutron therapy given in high-dose fractionation regimen resulted in complete tumor response compared to the standard photon therapy.

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32

Ekendahl, Daniela, Peter Rubovič, Pavel Žlebčík, Ivan Hupka, Ondřej Huml, Věra Bečková, and Helena Malá. "NEUTRON DOSE ASSESSMENT USING SAMPLES OF HUMAN BLOOD AND HAIR." Radiation Protection Dosimetry 186, no. 2-3 (November 7, 2019): 202–5. http://dx.doi.org/10.1093/rpd/ncz202.

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Abstract The unique feature of nuclear accidents with neutron exposure is the induced radioactivity in body tissues. For dosimetry purposes, the most important stable isotopes occurring in human body, which can be activated by neutrons, are 23 Na and 32 S. The respective activation reactions are as follows:23Na(n,γ)24Na and32S(n,p)32P. While sodium occurs in human blood, sulfur is present in human hair. In order to verify the practical feasibility of this dosimetry technique in conditions of our laboratory, samples of human blood and hair were irradiated in a channel of a training reactor VR-1.24Na activity was measured by gamma-ray spectrometry.32P activity in hair was measured by means of a proportional counter. Based on neutron-spectrum calculation, relationships between neutron dose and induced activity were derived for both blood and hair.

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33

Matsumoto, Shinnosuke, and Shunsuke Yonai. "EVALUATION OF NEUTRON AMBIENT DOSE EQUIVALENT IN CARBON-ION RADIOTHERAPY WITH ENERGY SCANNING." Radiation Protection Dosimetry 191, no. 3 (September 2020): 310–18. http://dx.doi.org/10.1093/rpd/ncaa166.

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Abstract In carbon-ion radiotherapy (CIRT), secondary neutrons are produced by nuclear interactions in the beamline devices or patient. Herein, the characteristics of secondary neutrons in CIRT with energy scanning (ES) were evaluated. Neutron ambient dose equivalents (H*(10)) were measured using WENDI-II. The neutron energy spectrum was calculated using the Monte Carlo simulation. Measurement and calculation were performed under realistic case scenarios using maximum beam energies (Emax) of 290, 350 and 400 MeV u −1. Moreover, H*(10) in ES was compared with H*(10) in range-shifter scanning (RS) and hybrid scanning (HS). H*(10) in Emax = 290 MeV u−1 was 65% less than that in Emax = 400 MeV u−1. At Emax = 350 MeV u−1, H*(10) in ES at θ = 120 was 42% of that at θ = 60. The neutron dose in ES CIRT decreased to approximately 60 and 70% of that in RS and HS CIRT, respectively, at 50-cm distance from the beam axis.

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Leite, A. M. M., M. G. Ronga, M. Giorgi, Y. Ristic, Y. Perrot, F. Trompier, Y. Prezado, G. Créhange, and L. De Marzi. "Secondary neutron dose contribution from pencil beam scanning, scattered and spatially fractionated proton therapy." Physics in Medicine & Biology 66, no. 22 (November 21, 2021): 225010. http://dx.doi.org/10.1088/1361-6560/ac3209.

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Abstract The Orsay Proton therapy Center (ICPO) has a long history of intracranial radiotherapy using both double scattering (DS) and pencil beam scanning (PBS) techniques, and is actively investigating a promising modality of spatially fractionated radiotherapy using proton minibeams (pMBRT). This work provides a comprehensive comparison of the organ-specific secondary neutron dose due to each of these treatment modalities, assessed using Monte Carlo (MC) algorithms and measurements. A MC model of a universal nozzle was benchmarked by comparing the neutron ambient dose equivalent, H*(10), in the gantry room with measurements obtained using a WENDI-II counter. The secondary neutron dose was evaluated for clinically relevant intracranial treatments of patients of different ages, in which secondary neutron doses were scored in anthropomorphic phantoms merged with the patients’ images. The MC calculated H*(10) values showed a reasonable agreement with the measurements and followed the expected tendency, in which PBS yields the lowest dose, followed by pMBRT and DS. Our results for intracranial treatments show that pMBRT yielded a higher secondary neutron dose for organs closer to the target volume, while organs situated furthest from the target volume received a greater quantity of neutrons from the passive scattering beam line. To the best of our knowledge, this is the first study to compare MC secondary neutron dose estimates in clinical treatments between these various proton therapy modalities and to realistically quantify the secondary neutron dose contribution of clinical pMBRT treatments. The method established in this study will enable epidemiological studies of the long-term effects of intracranial treatments at ICPO, notably radiation-induced second malignancies.

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35

Pradana, Luqman Satria, Utari Utari, Suharyana Suharyana, and Azizul Khakim. "ESTIMATION OF NEUTRON AND PROMPT PHOTON DOSE RATE DISTRIBUTION IN TMSR-500 USING MCNP6." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 24, no. 3 (November 9, 2022): 107. http://dx.doi.org/10.17146/tdm.2022.24.3.6692.

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Thorium Molten Salt Reactor-500 (TMSR-500), one of the Generation IV nuclear reactors, is designed by Thorcon International, Pte. Ltd, which is projected to be built in Indonesia. The reactor core is radially surrounded by B4C shielding, but not the upper part. As the silo hall sits above the reactor core and is accessible by reactor personnel, the dose rate must be calculated in the area to ensure the workers receive an annual dose below the acceptable limit. The dose rate from neutrons and photons as the result of fission reactions are the only sources to be calculated in this research, without taking the source from fission products into account. This research aims to obtain the dose rate distribution of neutrons and prompt photons using Monte Carlo code MCNP6. The reactor was assumed to operate at a nominal thermal power of 557 MWth. Dose rate calculation was obtained from flux Tally F4 and converted into dose rate using Dose Energy Dose Function (DEDF) factor. Conversion factors of flux to the dose were based on ICRP-21 and ANSI/ANS-6.1.1 1977. The result of the calculations showed that the distribution of neutron and prompt photon fluxes does not reach the silo hall.

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Fragopoulou, M., V. Konstantakos, M. Zamani, S. Siskos, T. Laopoulos, and G. Sarrabayrouse. "High sensitive depleted MOSFET-based neutron dosimetry." HNPS Proceedings 18 (November 23, 2019): 145. http://dx.doi.org/10.12681/hnps.2562.

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A new dosemeter based on a depleted Metal-Oxide-Semiconductor field effect transistor, sensitive to both neutrons and gamma radiation was manufactured at LAAS-CNRS Laboratory, Toulouse France. In order to be used for neutron dosimetry a thin film of lithium fluoride was deposited on the surface of the gate of the device. The characteristics of the dosemeter such as its response to neutron dose were investigated. The response in thermal neutrons was found to be high. In fast neutrons the response was lower than that of thermal neutrons but higher than the one presented in literature.

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Dorrah, Mahmoud E. "Paradoxical Effect of Neutron Shields on Neutron Dose from LINACs." International Journal of Scientific and Research Publications (IJSRP) 9, no. 5 (May 6, 2019): p8937. http://dx.doi.org/10.29322/ijsrp.9.05.2019.p8937.

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Metwally, Walid A., Osama A. Taqatqa, Mohammed M. Ballaith, Allan X. Chen, and Melvin A. Piestrup. "NEUTRON AND PHOTON DOSE MAPPING OF A DD NEUTRON GENERATOR." Radiation Protection Dosimetry 176, no. 3 (February 16, 2017): 258–63. http://dx.doi.org/10.1093/rpd/ncx004.

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Лисин and V. Lisin. "Evaluation of Therapeutic Gain \factor in Neutron Therapy Based on the Linear Quadratic Model." Medical Radiology and radiation safety 62, no. 1 (February 26, 2017): 65–70. http://dx.doi.org/10.12737/25063.

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Purpose: To study the dependencies of therapeutic gain factor (TGF) on dose of cyclotron-produced fast neutron beams using the linear-quadratic model (LQM) parameters characterizing radiation response in tumor and normal tissues. Material and methods: The TGF in neutron therapy was calculated as the ratio of the relative biological effectiveness of neutrons for tumor (RBE tumor) to relative biological effectiveness for normal tissue (RBE normal tissue). The LQM was used to calculate the dependencies of neutron RBE on the dose and therapeutic gain factor. We considered two cases: 1) neutron therapy for 3 types of tumors with different radiation response, where the same normal tissue was critical; 2) neutron therapy for the same tumor, when 3 types of normal tissues were taken as critical. Results: Based on calculations and analysis of published data, the dependencies of neutron RBE on dose for selected types of tumors and normal tissues were obtained. The following variants were considered: 1) RBE tumor > RBE normal tissue; 2) RBE tumor < RBE normal tissue, in both two variants, the dependen- cies in the therapeutic dose rate were convergent; 3) the dependencies of RBE tumor and RBE normal tissue on dose are crossed. The dependencies of TGF for neutron therapy on single boost doses and quantitative ratios between the LQM parameters characterizing radiation response of tumor and normal tissues were found. A multivariate ratio between the dependencies on dose of RBE tumor and RBE normal tissue was the cause of variety in the dependencies of TGF on dose. In the first case, the TGF increased with increasing (α/β)γ ratio and decreasing single dose, and the maximum value of TGF was equal to ~ 1.4. In the second case, TGF was < 1, i.e. the effectiveness of neutron therapy was lower than the effectiveness of gamma irradiation, but it was increased with higher single dose and lower radiosensitivity of normal tissue. In the third case, the dose at the intersection point (Di) was the boundary, and TGF was > 1 to the left of the boundary, and TPV was <1 to the right of the boundary, provided that D <Di, RBEtumor > OBEnormal tissue. Conclusion: The obtained results with known parameters of the LQM for tumor and normal tissues allowed us to make an appropriate choice between neutron and gamma- ray therapy in order to increase the effectiveness of treatment for cancer patients. It was shown that in the case of neutron therapy, the analysis of dependencies of TGF on dose allowed the optimal dose fractionation regimen to be selected.

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Rakotovao, L. O., Sidik Permana, R. H. Oetami, and Rasito. "Simulation of Neutron and Gamma Dose Rate of The TRIGA 2000 Reactor Using Monte Carlo Method." Journal of Physics: Conference Series 2328, no. 1 (August 1, 2022): 012007. http://dx.doi.org/10.1088/1742-6596/2328/1/012007.

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Abstract The neutron and gamma radiation doses were calculated from the operation of the 1 MW TRIGA 2000 Reactor in a simulation using the Monte Carlo method with MCNPX and PHITS program. Simulation is done by modelling the geometry of the reactor component materials and running it on a computer. The radiation source in the form of a fission reaction in the reactor core has been simulated using MCNPX to produce a dose of neutron and gamma radiation in the TRIGA 2000 core. Attenuation of neutron and gamma radiation by the reactor building is simulated using the PHITS code so that the neutron and gamma dose rates are obtained on the source, y, and z from reactor core. Interpolation of dose rate curves on large material thicknesses was carried out with the TVL neutron and gamma values of the simulation results for each reactor material. The simulation results shows that the gamma neutron dose rate outside the TRIGA 2000 reactor building is still below the dose limit value for radiation workers.

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Sasaki, Masao S., Satoru Endo, Masaharu Hoshi, and Taisei Nomura. "Neutron relative biological effectiveness in Hiroshima and Nagasaki atomic bomb survivors: a critical review." Journal of Radiation Research 57, no. 6 (September 2016): 583–95. http://dx.doi.org/10.1093/jrr/rrw079.

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Abstract The calculated risk of cancer in humans due to radiation exposure is based primarily on long-term follow-up studies, e.g. the life-span study (LSS) on atomic bomb (A-bomb) survivors in Hiroshima and Nagasaki. Since A-bomb radiation consists of a mixture of γ-rays and neutrons, it is essential that the relative biological effectiveness (RBE) of neutrons is adequately evaluated if a study is to serve as a reference for cancer risk. However, the relatively small neutron component hampered the direct estimation of RBE in LSS data. To circumvent this problem, several strategies have been attempted, including dose-independent constant RBE, dose-dependent variable RBE, and dependence on the degrees of dominance of intermingled γ-rays. By surveying the available literature, we tested the chromosomal RBE of neutrons as the biological endpoint for its equivalence to the microdosimetric quantities obtained using a tissue-equivalent proportional counter (TEPC) in various neutron fields. The radiation weighting factor, or quality factor, Qn, of neutrons as expressed in terms of the energy dependence of the maximum RBE, RBEm, was consistent with that predicted by the TEPC data, indicating that the chromosomally measured RBE was independent of the magnitude of coexisting γ-rays. The obtained neutron RBE, which varied with neutron dose, was confirmed to be the most adequate RBE system in terms of agreement with the cancer incidence in A-bomb survivors, using chromosome aberrations as surrogate markers. With this RBE system, the cancer risk in A-bomb survivors as expressed in unit dose of reference radiation is equally compatible with Hiroshima and Nagasaki cities, and may be potentially applicable in other cases of human radiation exposure.

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Bedenko, Sergey V., Vladimir V. Knyshev, Mariya Ye Kuznetsova, Igor O. Lutsik, and Igor V. Shamanin. "Peculiarities of the radiation formation in dispersed microencapsulated nuclear fuel." Nuclear Energy and Technology 5, no. 1 (March 20, 2019): 23–29. http://dx.doi.org/10.3897/nucet.5.33978.

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A computational study has been performed for various options of the thorium reactor core loading. Neutronic studies of fuel have been conducted, its isotopic composition has been calculated, and the alpha emitters and the sources of neutron and photon radiation in the microencapsulated nuclear fuel have been analyzed. The studies had the purpose of developing the methodology used to estimate the radiation characteristics of nuclear fuel with a complex inner structure. Emphasis is placed on calculating the quantitative and spectral composition of the neutrons formed as the result of (a, n) reactions on small- and average-mass nuclei. The ratio of the quantity of the neutrons resulting from the (a, n) reactions to the quantity of the neutrons formed as the result of spontaneous fission has been calculated for fuel with heterogeneous and homogeneous arrangements of fissionable and structural elements. The developed tools will make it possible to estimate the neutron radiation dose, to revise the traditional fresh and spent fuel handling procedures, and to estimate, using the Rossi alpha method, the neutron multiplication factor in deeply subcritical systems. The neutron yield and spectrum were calculated using an analytical model and verified codes such as WIMS-D5B, ORIGEN-APP, SOURCES-4C and SRIM-2013.

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Uhlář, Radim, and Petr Alexa. "MCNP APPROACHES FOR DOSE RATES MODELING IN LABORATORY FOR NEUTRON ACTIVATION ANALYSIS AND GAMMA SPECTROMETRY AT OSTRAVA." Radiation Protection Dosimetry 185, no. 1 (December 1, 2018): 116–23. http://dx.doi.org/10.1093/rpd/ncy209.

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Abstract The Laboratory for Neutron Activation Analysis and Gamma Spectrometry at the VŠB-Technical University of Ostrava was equipped with the neutron generator MP320 operating on the principle of the deuterium–tritium fusion and producing 108 neutrons/s at maximum. To ensure radiation protection of radiation workers and public outside the laboratory, the concrete shielding was designed and its protection efficiency was validated by MCNP simulations. Three approaches to calculate the dose rates were compared. The dose rates were estimated for the ORNL MIRD phantom located at the relevant positions (Tally F6 and *F8) and using the MCNPX mesh tally feature with the new ICRP Publication 116 flux-to-dose conversion factors. It was proven that the Approach II in which the absorbed dose rates due to neutrons for all organs are computed using the cell tally F6 and the photon dose calculation is performed by the *F8 energy deposition tally is the most valuable one.

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Zorloni, Gabriele, Geert Bosmans, Thomas Brall, Marco Caresana, Marijke De Saint-Hubert, Carles Domingo, Christian Ferrante, et al. "Joint EURADOS WG9-WG11 rem-counter intercomparison in a Mevion S250i proton therapy facility with Hyperscan pulsed synchrocyclotron." Physics in Medicine & Biology 67, no. 7 (March 28, 2022): 075005. http://dx.doi.org/10.1088/1361-6560/ac5b9c.

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Abstract Objective. Proton therapy is gaining popularity because of the improved dose delivery over conventional radiation therapy. The secondary dose to healthy tissues is dominated by secondary neutrons. Commercial rem-counters are valuable instruments for the on-line assessment of neutron ambient dose equivalent (H*(10)). In general, however, a priori knowledge of the type of facility and of the radiation field is required for the proper choice of any survey meter. The novel Mevion S250i Hyperscan synchrocyclotron mounts the accelerator directly on the gantry. It provides a scanned 227 MeV proton beam, delivered in pulses with a pulse width of 10 μs at 750 Hz frequency, which is afterwards degraded in energy by a range shifter modulator system. This environment is particularly challenging for commercial rem-counters; therefore, we tested the reliability of some of the most widespread rem-counters to understand their limits in the Mevion S250i stray neutron field. Approach. This work, promoted by the European Radiation Dosimetry Group (EURADOS), describes a rem-counter intercomparison at the Maastro Proton Therapy centre in the Netherlands, which houses the novel Mevion S250i Hyperscan system. Several rem-counters were employed in the intercomparison (LUPIN, LINUS, WENDI-II, LB6411, NM2B-458, NM2B-495Pb), which included simulation of a patient treatment protocol employing a water tank phantom. The outcomes of the experiment were compared with models and data from the literature. Main results. We found that only the LUPIN allowed for a correct assessment of H*(10) within a 20% uncertainty. All other rem-counters underestimated the reference H*(10) by factors from 2 to more than 10, depending on the detector model and on the neutron dose per pulse. In pulsed fields, the neutron dose per pulse is a fundamental parameter, while the average neutron dose rate is a secondary quantity. An average 150–200 μSv/GyRBE neutron H*(10) at various positions around the phantom and at distances between 186 cm and 300 cm from it was measured per unit therapeutic dose delivered to the target. Significance. Our results are partially in line with results obtained at similar Mevion facilities employing passive energy modulation. Comparisons with facilities employing active energy modulation confirmed that the neutron H*(10) can increase up to more than a factor of 10 when passive energy modulation is employed. The challenging environment of the Mevion stray neutron field requires the use of specific rem-counters sensitive to high-energy neutrons (up to a few hundred MeV) and specifically designed to withstand pulsed neutron fields.

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Razghandi, S., K. Karimi-Shahri, and M. M. Firoozabadi. "Evaluation of neutron spectra and dose equivalent from a Varian 2100C/D Medical Linear Accelerator: Monte Carlo simulation and a literature review." Radioprotection 56, no. 2 (March 11, 2021): 93–101. http://dx.doi.org/10.1051/radiopro/2021002.

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In this study was carried out a review according to experimental and Monte Carlo studies in the literature on the neutron production from 18 MV, Varian 2100C/D linac. The effects of these neutrons were investigated on the total fluence, the energy spectra, and the dose equivalent. These factors were calculated as a function of depth and the radiation field size by simulation of linac head using of MCNPX2.6.0 code. The neutron strength was found equal to 1.23 × 1012 nGy−1.The results showed that with increasing the field size from 5 × 5 to 40 × 40 cm2, the neutron fluence and dose equivalent in the water phantom rose to the maximum value for 25 × 25 cm2 field (3.05 × 107 ncm−2Gy−1 and 3.14 mSvGy−1 respectively) and then decreased with increasing the field size. According to the results, the magnetite-steel, ordinary, and limonite-steel concrete walls significantly increased the neutron dose equivalent for about 27.4%, 17.2%, and 13.5%, respectively.

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YAMAGUCHI, Yasuhiro. "Effective Dose for External Neutron Exposure." RADIOISOTOPES 42, no. 1 (1993): 35–36. http://dx.doi.org/10.3769/radioisotopes.42.35.

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Fragopoulou, M., S. Stoulos, M. Manolopoulou, M. Krivopustov, and M. Zamani. "Dose measurements around spallation neutron sources." Radiation Protection Dosimetry 132, no. 3 (October 28, 2008): 277–82. http://dx.doi.org/10.1093/rpd/ncn280.

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Shonka, Joseph. "Schauer Neglects Neutron Dose in Aviation." Health Physics 114, no. 4 (April 2018): 461. http://dx.doi.org/10.1097/hp.0000000000000832.

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Veinot, K. G., N. E. Hertel, M. M. Hiller, and K. F. Eckerman. "Neutron dose coefficients for local skin." Journal of Radiological Protection 40, no. 2 (May 13, 2020): 554–82. http://dx.doi.org/10.1088/1361-6498/ab805e.

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Vondracek, V., M. Kralik, K. Turek, and M. Navratil. "NEUTRON DOSE WITH 18 MV IMRT." Radiotherapy and Oncology 92 (August 2009): S189. http://dx.doi.org/10.1016/s0167-8140(12)73087-6.

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Journal articles on the topic 'Neutron dose'

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