Fast Neutron Detection in Nuclear Material Photofission Assay Using a 15 MeV Linear Electron Accelerator

The purpose of this research was to use a 15 MeV (K15 model by Varian) linear electron accelerator (linac) for the photon assay of special nuclear materials (SNM). First, the properties of the photon radiation probe were determined. The stochastic radiation transport code, MCNP5, was used to develop computational models for the linac. The spectral distribution of photons as well as dose rate contour maps of the UNLV accelerator facility were computed for several linac operating configurations. These computational models were validated through comparison with experimental measurements of dose rates. The linac model was used to simulate the photon interrogation of SNM targets of various compositions and shielding materials. The spectra of neutrons produced by the irradiation of shielded SNM was characterized. The effects of shielding material and the SNM enrichment on the neutron yields following photon assay were determined. It was determined that the radiation signatures following the photon assay of SNM consisted of photons and neutrons produced from the fissions, in addition to neutrons produced from photonuclear reactions. The EJ-299-33A plastic scintillator was evaluated for this study due to its ability to discriminate between fast neutrons and gamma rays. The neutron coincidence measurement option was also evaluated. The detector response functions were determined for different incident neutron energies. Further, it was computationally shown that an array of EJ-299-33A detectors allows to measure neutron multiplicity, enabling discrimination between fission neutrons and the photoneutrons

Thermal and epithermal neutrons in the vicinity of the Primus Siemens biomedical accelerator

In this paper, the thermal and epithermal neutron fluence distributions in the vicinity of the Primus Siemens accelerator are presented. The measurements were carried out by the use of the neutron activation method for 15 MV X-rays and electron beams of 18 MeV and 21 MeV. From the radiation safety point of view for the hospital personnel, it is important to know the thermal and epithermal neutron fluence distribution in the vicinity of the accelerator because the neutrons interacting with atoms of a medium by various processes induce the activity of objects (accelerator, other apparatus etc.) and walls in the treatment room. The thermal and epithermal neutron capture, particularly, in high atomic number materials of the accelerator head can be a significant source of gamma radiation and it has to be taken into account for estimation of the work safety of the personnel. Values of the neutron fluence were normalized to the maximum photon (or electron) dose Dmax,γ (e) measured at the central axis of therapeutic X-ray (or electron) beam in a water phantom. The thermal neutron fluences measured during the 15 MV X-ray emission varied between 1.1 × 105 n · cm−2· Gy−1 and 4.4 × 105 n · cm−2· Gy−1 whereas the epithermal neutron fluences ranged from 0.2 × 105 n · cm−2· Gy−1 to 1.8 × 105 n · cm−2· Gy−1. In the case of electron beams, the neutron fluence measurements were performed only at the isocentre. The obtained thermal and epithermal neutron fluences were 1.2 × 104 n · cm−2· Gy−1 and 0.6 × 104 n · cm−2· Gy−1, respectively, for the 18 MeV electrons. In the the case of the 21 MeV electron beams the thermal neutron fluence was 2.0 × 104 n · cm−2· Gy−1 whereas the epithermal neutron fluence was 0.8 × 104 n · cm−2· Gy−1

Radiation Therapy and Dosing Material Transport Methodology

A technique is examined here that utilizes high energy beta decays from a short lived radioisotope to treat medical conditions such as shallow cancerous lesions. A major benefit of beta particle interaction in tissue is a fixed penetration depth for the charged particle, with dose limited to the ultimate range of the beta particle. This method improves on some current techniques of radioactive brachytherapy, where seeds are placed inside patients through temporary or permanent implantation in order to kill cancerous cells or inhibit growth of tissue. The use of low energy gamma-rays is the most common method of treatment currently for brachytherapy, with Ir-192 used in most high dose rate procedures. The 73.8 day half-life of Ir-192 means frequent replacement and the requirement to deal with the logistics of constantly decaying, fixed radioactive sources. This method instead utilizes the short 14.1 second half-life of indium-116 to quickly deliver dose to a treatment area while decaying to a stable ground state. Since the isotope is very short lived, pumping is used to transport the isotope in a room temperature liquid eutectic between the “activation site” where the In-116 is created, and the “application site” where it is allowed to decay over a target area. In-116 is produced at the activation site through neutron capture on the stable isotope In-115. This radioactive In-116 is then pumped through a sealed, closed loop system using a peristaltic to an application site where it is allowed to decay for one minute, enough time to pass through over four 14.1 second half-lives and reach a stable ground state. This applicator is a sealed spreader surface with a thin barrier to allow passage of the decay betas. The loop is then repeated with the In-115 activated again and pumped. This work examined the feasibility of this method with three types of neutron sources including a fixed Pu-Be source, a Dense Plasma Focus pulsed fusion neutron source and an XRay producing accelerator using photoneutrons from a beryllium target. Radiation transport modeling was used to determine the efficacy of this technique on two higher output neutron sources, including the use of a standard clinical accelerator used for external beam therapy, the Varian Clinac 2200C. Neutron output from the Clinac was modeled based on photonuclear production in the accelerator components when operated at 20MeV. Dose outputs were found to be viable for clinical use when the system is used on a Clinac due to the substantial photoneutron output. In addition to therapy, this work demonstrated the ability to measure neutron fluence at a remote location by measuring decaying In-116 from the eutectic after irradiation. In particular, this was demonstrated with the pulsed DPF source and compared with existing yield measurement techniques. The yield of neutrons from pulsed sources can be difficult to measure due to the intense and brief burst, preventing the use of normal radiation detectors that measure radiation over time. Pulsed sources instead require activation of materials to create a signature of the yield magnitude which is then counted after the pulse. This method showed excellent agreement with an existing method of beryllium activation detection

Aspetti avanzati di radioprotezione nell'uso di acceleratori di particelle in campo medico

In this work, the well-known MC code FLUKA was used to simulate the GE PETrace cyclotron (16.5 MeV) installed at “S. Orsola-Malpighi” University Hospital (Bologna, IT) and routinely used in the production of positron emitting radionuclides. Simulations yielded estimates of various quantities of interest, including: the effective dose distribution around the equipment; the effective number of neutron produced per incident proton and their spectral distribution; the activation of the structure of the cyclotron and the vault walls; the activation of the ambient air, in particular the production of 41Ar, the assessment of the saturation yield of radionuclides used in nuclear medicine. The simulations were validated against experimental measurements in terms of physical and transport parameters to be used at the energy range of interest in the medical field. The validated model was also extensively used in several practical applications uncluding the direct cyclotron production of non-standard radionuclides such as 99mTc, the production of medical radionuclides at TRIUMF (Vancouver, CA) TR13 cyclotron (13 MeV), the complete design of the new PET facility of “Sacro Cuore – Don Calabria” Hospital (Negrar, IT), including the ACSI TR19 (19 MeV) cyclotron, the dose field around the energy selection system (degrader) of a proton therapy cyclotron, the design of plug-doors for a new cyclotron facility, in which a 70 MeV cyclotron will be installed, and the partial decommissioning of a PET facility, including the replacement of a Scanditronix MC17 cyclotron with a new TR19 cyclotron.In questo lavoro, il codice Monte Carlo (MC) FLUKA è stato utilizzato per simulare il ciclotrone GE PETtrace (16.5 MeV) installato presso l’azienda ospedaliera “S. Orsola-Malpighi” (Bologna, IT), quotidianamente utilizzato per la produzione di radiofarmaci PET. Le simulazioni sono state effettuate per valutare diversi fenomeni e quantità d’interesse radiologico tra cui l’equivalente di dose ambientale nell’intorno dell’acceleratore, il numero di neutroni emessi per protone incidente e la loro distribuzione spettrale, l’attivazione dei componenti del ciclotrone e delle pareti del bunker, l’attivazione dell’aria interna al bunker ed in particolare la produzione di 41Ar, la resa a saturazione di radionuclidi d’interesse in medicina nucleare. Le simulazioni sono state validate, in termini di parametri fisici e di trasporto da utilizzare nel range energetico caratteristico delle applicazioni mediche, con una serie di misure sperimentali. Il modello MC validato è stato quindi applicato ad altri casi pratici quali lo studio di fattibilità della produzione diretta in ciclotrone di 99mTc, la produzione di radionuclidi ad uso medico con il ciclotrone TR13 (13 MeV) installato presso il centro di ricerca TRIUMF (Vancouver, CA), la progettazione completa del nuovo centro PET dell’ospedale “Sacro Cuore-Don Calabria” di Negrar (Verona, IT), incluso il ciclotrone ACSI TR19 (19 MeV), lo studio del campo di dose nell’intorno di un sistema di selezione dell’energia (degrader) di un ciclotrone per terapia, la progettazione di specifiche “porte a tappo” per un sito di produzione di radionuclidi ad uso medico, in cui verrà installato un ciclotrone da 70 MeV e sei diverse beam line, e per il parziale decommissioning di un centro PET e la sostituzione di un ciclotrone Scanditronix MC17 (17 MeV), attualmente installato, con una nuova unità TR19

Measurements of Neutron Activation and Dose Rate Induced by High-Energy Medical Linear Accelerator

Purpose: During the treatments of cancer patients with a linear accelerator (LINAC) using photon beams with energies ≥8 MV, the components inside the LINAC head get activated through the interaction of photonuclear reaction (γ, n) and neutron capture (n, γ). We used spectroscopy and measured the dose rate for the LINAC in operation after the treatment ended. Methods: We performed spectroscopy and dose rate measurements for three units of LINACs with a portable high-purity Germanium (HPGe) detector and a survey meter. The spectra were obtained after the beams were turned off. Spectroscopy was conducted for 3,600 seconds, and the dose rate was measured three times. We identified the radionuclides for each LINAC. Results: According to gamma spectroscopy results, most of the nuclides were short-lived radionuclides with half-lives of 100 days, except for 60Co, 65Zn, and 181W nuclides. The dose rate for three LINACs obtained immediately in front of the crosshair was in the range of 0.113 to 0.129 µSv/h. The maximum and minimum dose rates measured on weekends were 0.097 µSv/h and 0.092 µSv/h, respectively. Compared with the differences in weekday data, there was no significant difference between the data measured on Saturday and Sunday. Conclusions: Most of the detected radionuclides had half-lives <100 days, and the dose rate decreased rapidly. For equipment that primarily used energies ≤10 MV, when the equipment was transferred after at least 10 minutes after shutting it down, it is expected that there will be little effect on the workers’ exposure.ope

Thermal Neutron Relative Biological Effectiveness Factors for Boron Neutron Capture Therapy from In Vitro Irradiations

The experimental determination of the relative biological effectiveness of thermal neutron factors is fundamental in Boron Neutron Capture Therapy. The present values have been obtained while using mixed beams that consist of both neutrons and photons of various energies. A common weighting factor has been used for both thermal and fast neutron doses, although such an approach has been questioned. At the nuclear reactor of the Institut Laue-Langevin a pure low-energy neutron beam has been used to determine thermal neutron relative biological effectiveness factors. Different cancer cell lines, which correspond to glioblastoma, melanoma, and head and neck squamous cell carcinoma, and non-tumor cell lines (lung fibroblast and embryonic kidney), have been irradiated while using an experimental arrangement designed to minimize neutron-induced secondary gamma radiation. Additionally, the cells were irradiated with photons at a medical linear accelerator, providing reference data for comparison with that from neutron irradiation. The survival and proliferation were studied after irradiation, yielding the Relative Biological Effectiveness that corresponds to the damage of thermal neutrons for the different tissue types.Asociacion Espanola Contra el Cancer (AECC) PS16163811PORRSpanish MINECO FIS2015-69941-C2-1-PJunta de Andalucia P11-FQM-8229Campus of International Excellence BioTic P-BS-64University of Granada Chair Neutrons for Medicine: the Spanish Fundacion ACSAsociacion Capitan AntonioFundacion ACSLa Kuadrilla de IznallozSonriendo Se Puede Gana

The Quantitative Measurement of Neutron Induced Activity in Biomedical Applications.

The principles of In Vivo Neutron Activation Analysis, IVNAA, and the present status of IVNAA methods and other in vivo elemental composition techniques have been outlined. Description has been given of the modifications made to the IVNAA facility used in the present study and the subsequent effects on the performance of the system have been discussed. Detection limits, using the ’prompt’ neutron activation technique for the major body elements, sodium, chlorine and nitrogen were found to be 220 ppm, 140 ppm and 1.55% by weight, respectively. It was found that neither calcium nor phosphorus could be measured using the ’prompt’ technique at acceptable dose levels to the subject. It has been shown that measurement of body chlorine concentrations through the 37Cl (n,&gamma;) 38mCl reaction is not feasible using ’cyclic’ activation analysis at acceptable dose levels delivered to the subject. Determination of cadmium and selenium concentrations in a liver phantom was carried out, during the same experiment, using the technique of alternate ’prompt’ and ’cyclic’ activation analysis. This allows for the collection of the ’prompt’ gamma-ray data with no further dose delivered to the subject. Detection limits of 13 ppm and 5.8 ppm for Cd and Se were obtained, respectively. The origin of the interfering photopeak in 77mSe measurements was not conclusively identified but some of the possible sources have been outlined and suggestions have been made for further investigations. Neutron inflicted damage of germanium based semiconductor detectors has been discussed and a method of Ge(Li) crystal repair has been described that is expected to lead to full fast neutron damaged detector regeneration, provided that the necessary active outgassing of the detector vacuum enclosure is incorporated. A Monte Carlo aided Fortran-77 computer programme for the calculation of the average solid angle subtended by a collimated detector at the photon emitting source which addresses the collimator edge penetration was developed and tested. The strength of the simulation technique in providing pre-experimental information in a variety of investigations where collimated detectors are used for gamma-ray measurements has been demonstrated. The goal of performing IVNAA as an ’absolute’ method has been partially addressed; it has been shown that the programme can be used to: (i) determine the volume of the activated target viewed by the collimated detector and (ii) to estimate the effect of the neutron flux non-uniformity within the activated volume of interest. Finally, the photoneutron field around a medical electron accelerator has been determined using ’bare’ activation detectors. A review has been given of the past photoneutron measurements and the results of the present study have been discussed. Evidence of photoneutron production within the patient body has been presented. The ’in beam’ photoneutron dose equivalent contribution on the patient plane was found to be 8.7 +- 30% mSv per Gy of photons

Measurement of Neutron Activation from a High Energy Varian Linear Accelerator

Linear accelerators producing photons above 10 MeV may induce photonuclear reactions in high Z components of the accelerator. These liberated neutrons can then activate the structural components of the accelerator and other materials in the beam path through neutron capture reactions. The induced activity within the accelerator may contribute to additional dose to both patients and personnel. This project seeks to determine the total activity and activity per activated isotope following irradiation from a Varian medical linear accelerator at energies above 10 MeV. A Varian 21iX accelerator was used to irradiate a 30 cm x 30 cm x 20 cm solid water phantom with 15 MV x-rays. The phantom was placed at a source-to- surface distance (SSD) of 100 cm and at the center of a 20 cm x 20 cm field. Activation induced gamma spectra were acquired over a 5 minute interval after 1 and 15 minutes from completion of the irradiation. All measurements were made using a CANBERRA Falcon 5000 Portable high purity germanium (HPGe) detector. The majority of measurements were made in scattering geometry with the detector situated at 90° to the incident beam, 30 cm from the side of the phantom and approximately 10 cm from the top. A 5 minute background count was acquired and automatically subtracted from all subsequent measurements. Photon spectra were acquired for both open and MLC fields and activities for each nuclide were estimated from detector efficiencies as determined from Monte Carlo simulations. Based on spectral signatures, the following nuclides were identified: 56 Mn, 62 Cu, 64 Cu, 82 Br, 99 Mo, 122 Sb, 124 Sb, and 187 W. In all cases, estimated activities from the activation products were in the microcurie range