Determination of Monitor Doses in Neutron Therapy Using the U-120 Cyclotron

Purpose Analyze the various methods for determining the monitor doses in neutron therapy using the U-120 cyclotron and to choose the monitoring method that provides the highest accuracy in dose delivery to the tumor. Material and methods The distributions of the absorbed dose of the therapeutic beam from the U-120 cyclotron were measured in a tissue-equivalent medium using the differential method, in which two ionization chambers with different sensitivity to neutron radiation were used. A comparison of radiation effects on tissues using various techniques of determining the monitor doses was made. The linear-quadratic model was used to assess responses to ionizing radiation. Results Dosimetry studies revealed that the therapeutic beam of the U-120 cyclotron contains concomitant gamma radiation, the contribution of which to the total neutron-photon dose increases with increasing depth of the irradiated medium. The presence of gamma radiation in the neutron beam dictate the need to find the correct method for monitoring neutron therapy. A comparison of radiation effects on the tumor tissue using different techniques of determining the monitor doses was made. It was found that at equal neutron-photon doses, the neutron dose in the tumor changed depending on its depth. It can lead to an incorrect conclusion about the effectiveness of neutron therapy depending on a single dose as well as in relation to various dose fractionation schedules. Conclusion The analysis of the results obtained showed that the problem can be most accurately solved using a technique in which the monitor coefficient and monitor doses are determined from the distribution of the neutron dose, taking into account the contribution of the gamma radiation dose to the total neutron-photon dose.

Kajian pemantauan dosis neutron terhadap pekerja radiasi pada pengoperasian Linac dengan energi foton ≥ 10 MV

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.

Measurement of Neutron Dose Equivalent within and Outside of a LINAC Treatment Vault Using a Neutron Survey Meter

This work concerns neutron doses associated with the use of a Siemens Primus M5497 electron accelerator, which is operated in the photon mode at 15 MV. The conditions offer a situation within which a fraction of the bremsstrahlung emission energies exceed the photoneutron threshold. For different field sizes, an investigation has been made of neutron dose equivalent values at various measurement locations, including: (i) At the treatment table, at a source-surface distance of 100 cm; (ii) at the level of the floor directly adjacent to the treatment table; and (iii) in the control room and patient waiting area. The evaluated neutron dose equivalent was found to range from 0.0001 to 8.6 mSv/h, notably with the greatest value at the level of the floor directly adjacent to the treatment couch (8.6 mSv/h) exceeding the greatest value on the treatment table (5.5 mSv/h). Low values ranging from unobservable to between 0.0001 to 0.0002 mSv/h neutron dose were recorded around the control room and patient waiting area. For measurements on the floor, the study showed the dose equivalent to be greatest with the jaws closed. These data, most particularly concerning neutron distribution within the treatment room, are of great importance in making steps towards improving patient safety via the provision of protective measures.

Japanese pediatric and adult atomic bomb survivor dosimetry: potential improvements using the J45 phantom series and modern Monte Carlo transport

The radiation exposure estimates for the atomic bomb survivors at Hiroshima and Nagasaki have evolved over the past several decades, reflecting a constant strive by the Radiation Effects Research Foundation (RERF) to provide thorough dosimetry to their cohort. Recently, a working group has introduced a new series of anatomical models, called the J45 phantom series, which improves upon those currently used at RERF through greater age resolution, sex distinction, anatomical realism, and organ dose availability. To evaluate the potential dosimetry improvements that would arise from their use in an RERF Dosimetry System, organ doses in the J45 series are evaluated here using environmental fluence data for 20 generalized survivor scenarios pulled directly from the current dosimetry system. The energy- and angle-dependent gamma and neutron fluences were converted to a source term for use in MCNP6, a modern Monte Carlo radiation transport code. Overall, the updated phantom series would be expected to provide dose improvements to several important organs, including the active marrow, colon, and stomach wall (up to 20, 20, and 15% impact on total dose, respectively). The impacts were especially significant for neutron dose estimates (up to a two-fold difference) and within organs which were unavailable in the previous phantom series. These impacts were consistent across the 20 scenarios and are potentially even greater when biological effectiveness of the neutron dose component is considered. The entirety of the dosimetry results for all organs are available as supplementary data, providing confident justification for potential future DS workflows utilizing the J45 phantom series.