Neutron Dose Distribution Research Articles

Radiation environment of an MT-22 microtron

The authors present and analyze the results of full-scale measurements of the neutron and bremsstrahlung fields around the MT-22 microtron. Dose rates for neutron radiation and bremsstrahlung were measured both for the radiation emerging from the moderating target assembly impinged directly on the concrete shielding and for radiation emerging from the target assembly surrounded on three sides with lead shielding. The second case was found to provide the worst radiation protection; the moderating target was unprotected at the top while the lead itself served as a thick neutron converter. Bremsstrahlung was determined with a scintillation detector while neutron dose distributions were measured with a corona counter. Some activation reactions and their resulting radionuclides, generated in the irradiation of target and shielding materials, are given.

Shielding design and detection of neutrons from medical and industrial electron accelerators--simple method of design calculation for neutron shielding

The neutron leakage from medical and industrial electron accelerators has become an important problem and its detection and shielding is being performed in their facilities. This study provides a new simple method of design calculation for neutron shielding of those electron accelerator facilities by dividing into the following five categories; neutron dose distribution in the accelerator room, neutron attenuation through the wall and the door in the accelerator room, neutron and secondary photon dose distributions in the maze, neutron and secondary photon attenuation through the door at the end of the maze, neutron leakage outside the facility-skyshine.

A neutron-photon treatment planning intercomparison

In order to assess different treatment philosophies in different neutron therapy centres as well as to gauge the quality of neutron dose distributions relative to corresponding photon plans, a neutron-photon treatment planning intercomparison has been organized amongst the neutron therapy centres in Europe and the U.S.A. Treatment plans were compared for patients with a tumour in the floor of the mouth, a carcinoma of the base of the tongue and a large bladder carcinoma. A wide variety in treatment plans was received. This was mainly caused by the different assumptions on the treatment margin around the tumour and whether or not the lymph nodes should be irradiated electively. Within the same institute, however, similar types of plans were used for both the neutron and the photon irradiations. The homogeneity of the adsorbed dose over the target volume can be made almost the same for photon and neutron irradiations, although this may require a more complicated technique for neutron treatment. The relative absorbed dose to normal tissues was somewhat higher for the neutron treatment. There is an increased chance of the occurrence of a hot spot with the relatively low energy neutron beams compared to the photon beams.

Evaluation of Neutron Skyshine from a Cyclotron

The dose distribution and the spectrum variation of neutrons due to the skyshine effect have been measured with various detectors in the environment surrounding the cyclotron of the Institute for Nuclear Study, University of Tokyo. The source neutrons were produced by stopping a 52-MeV proton beam into a carbon beam stopper and were extracted upward from the opening in the concrete shield surrounding the cyclotron and then leaked into the atmosphere through the cyclotron building. The dose distribution and the spectrum of neutrons near the beam stopper were also measured in order to get information on the skyshine source. The measured skyshine neutron spectra and dose distribution were analyzed with two codes, MMCR2 and SKYSHINE-II, with the result that the calculated results are in good agreement with the experiment. Valuable characteristics of this experiment are the determination of the energy spectrum and dose distribution of source neutron and the measurement of skyshine neutrons from an actual large-scale accelerator building to the exclusion of direct neutrons transported through the air. This experiment must be useful as a kind of benchmark experiment on the skyshine phenomenon.

Calculation of Low-Energy Neutron Dose Indices and Depth Doses in the ICRU Tissue Sphere

Detailed neutron dose distributions were calculated for the ICRU tissue sphere for broad, parallel beams of incident neutrons with 10 different energies ranging from thermal to 0.3 MeV. The calculation was carried out by the Monte Carlo method on the CYBER 175 computer of the University of Illinois. From these calculated data, a set of fluence-to-dose-index conversion factors, needed to establish the allowable limits for exposure to external neutron beams, was obtained. Normalized depth doses along the principal axis at depths of 0.007, 0.3 and 1.0 cm are also provided. The fluence-to-dose-index conversion factors are compared to previous conversion factors based on cylindrical phantoms, and any differences are explained.

Radiation Protection Monitoring Around High Energy Proton Accelerators Using Thermoluminescence Dosemeters

Thermoluminescence detectors are in use at the European Organisation for Nuclear Research (CERN) for various applications in the field of radiation protection monitoring. A network of 6LiF and 7LiF detectors inside 12.5 cm 0 x 1.5 cm polyethylene cylinders monitors the spatial gamma and neutron dose distribution on the CERN sites and is used for natural radiation background measurements outside CERN at the houses of CERN employees. 6LiF and 7LiF detectors inside 25 cm diameter spherical neutron moderators are used for area monitoring. A system of CaF2:Dy TLDs is used for photon and muon dose monitoring. In addition, TLDs are in routine use for monitoring radioactive air releases in accelerator stacks. The application of 6LiF and 7LiF detectors for personal and accident dosimetry is also discussed.

Neutron dose distribution measurements in a water filled phantom by SSNTDs x

ABSTRACT Kodak Pathe LR115 II type cellulose nitrate track detectors with different (n, ∞) radiators were calibrated in various isotropic neutron fields and used for neutron dose distribution measurements in a water filled elliptical phantom irradiated with neutrons from different light water moderated reactors.

PION-I. A Monte Carlo Computer Program for Calculations with Negative Pion Beams

An earlier publication described a computer program for calculating depth-dose distributions in water phantons irradiated by positive or negative pion beams. This program, now called PION-I, has been substantially improved and modified as described here with emphasis on applications of negative pions to cancer radiotherapy. Modifications that have been made include a much improved treatment of multiple Coulomb scattering, the addition of range straggling, the addition of elastic scattering of pions from heavy nuclei, a new method of computing the neutron dose distribution around pion stars, provisions for calculating cell survival, and the inclusion of a code for computing isodata contours. Calculations with PION-I can also include the effects of inhomogeneities, such as regions of bone, lung, or air in tissue phantoms. A number of examples are given of the kind of information that can be obtained with PION-I. The relative effects of various physical processes on dose distribution patterns are also assessed.

Neutron beam characterization by means of a radioactivation image transfer (RIT) technique

Fast and thermal neutron distributions in phantom have been recorded by means of neutron induced radioactivity in aluminum and copper sheets, respectively. The induced β-decays in the metal sheets are used to form radiographic images of the fast and thermal neutron distributions on diagnostic X-ray film via autoradiography. Film calibration then translates the optical density map into the fast or slow neutron flux distribution in phantom; and, since the neutron spectrum does not change significantly with depth, the fast neutron flux distribution also maps the fast neutron dose distribution. The present radioactivation-image transfer (RIT) technique has several distinct advantages over ionization chamber measurements for neutron flux distributions: (1) the RIT method requires two thirds less beam time than ion-chamber measurements, (2) the induced activity is insensitive to -/-rays, (3) the equipment requisite for data acquisition is far less costly than an automated ionization chamber apparatus and (4) a permanent analog record is produced which can be examined at any later date.

Neutron dose distributions in tissue irradiated by negative pion beams

Monte Carlo computations are performed for four idealized uniform, circular negative pion beams. The individual dose contributions are shown from direct ionization by pions, from secondary protons, from neutrons, and from the heavier particles produced by nuclear absorption of pions. The resulting neutron dose distributions, which are important in therapeutic applications, are analyzed. Isodose contours for the total doses are presented together with isodose contours for neutron doses. The results show quantitative trends in the neutron dose distributions for different beam conditions and different treatment-volume sizes. They also show the sharpness of the boundaries of the high-dose regions that can be obtained with pion beams. The calculations indicate that only under special circumstances would the neutron dose outside the tumor volume limit the treatment dose.

Determination of Doses from Neutrons Based on Measuring Activities Induced in a Human Phantom Irradiated at a Nuclear Reactor

The whole-body neutron dose was obtained from the measurement of 21Na-activity in three parts of the phantom (head, torso, legs). The measurement of 32P-activity in samples of nails and hairs was used for determination of the neutron dose distribution. Doses in an interval of (0.1—0.5) J. kg−1 was determined with relative errors of 17.4% (from the 24Na-activity) and 23.3% (from the 32P-activity); in an interval of (0.01—0.11) J. Kg−1 with a relative error of 45% from the 24Na-activity. The approximation of the real spectrum by the “dE/E” spectrum in a neutron energy interval of 0.48 eV to 1 MeV and by the undisturbed fission spectrum above 0.1 MeV energy of neutrons results in errors of 69.8% for a dose based on the 24Na-activity in fast neutron region and of 28.8% for the same dose based on 32P-activity. The error of the dose based on the 24Na-activity will be of 44.9% in the region of intermediate neutrons.

Perturbation of neutron dose distributions by lung tissue.

The influence of lung tissue on depth dose distributions, produced by beams of fast neutrons (generated at the Naval Research Laboratory cyclotron by bombarding a thick Be target with 35‐MeV deuterons) was determined. Muscle‐equivalent ionization chambers were utilized in conjunction with thermoluminescent dosimeters to establish the neutron and gamma ray‐dose deposited in a phantom of the thoracic region of the body containing lung components. All irradiations were carried out at a target–skin distance of 125 cm with neutron beams of 10 × 10‐cm field size at the surface of the phantom. From these measurements, the authors conclude that the presence of lung tissue results in higher dose levels for points at depths greater than the midpoint of the lungs as a result of the increased transmission of neutrons by low‐density lung tissue. An examination of the dose measurements revealed that the portion of the total dose due to gamma rays is of the order of 4–10% for points in lung and muscle‐equivalent fluid.

Calculation of the HPRR neutron spectrum for simulated nuclear accident conditions.

Since 1964, the Health Physics Research Reactor (HPRR) at the Oak Ridge National Laboratory has been used to simulate nuclear excursions to allow intercomparisons of accident dosimetry systems. Generally, activation foils or threshold-activation foils are used to measure the neutron fiuence. Usually, data from these detectors are used to provide the neutron dose and the general shape of the neutron energy spectrum. However, some participants have used an iterative technique to calculate a spectral distribution to fit the experimental results. A calculational study was begun to provide the HPRR neutron spectrum for three typical accident exposure conditions. Results of these calculations are compared with spectra derived from the use of aeveral dosimeters under actual exposure conditions at the HPRR. Also included in the results of the calculation were neutron dose distributions as a function of distance from the reactor. These results are compared with the results of measurements performed during the intercomparison Study of 1971. (auth)

Neutron dose distributions for californium-252 sources

The neutron dose distribution in tissue-equivalent material for a 252Cf point source has been calculated by the Monte Carlo method. Using this data, the dose-rate distribution for an implantable 252Cf line source has been obtained, and the results are compared with available experimental data. The calculated neutron isodose curves for line sources of length 1·35 and 2 cm are also presented. Pacпpeдлeкйe дoэы кeйtpoHoB эквйвaлeHtHям мatrйaлe йctoч;плeHйe дoэы> HeфtroHoв в для toч;Hoгo йctoч;Hйкa 252Cf BыcчBaлocб мetoдoм MoHte Кarлo. Пrй йcпoлэoBaHйй эtйч дaHHыч былa пoычeHa мoшHotь дoэы pacпpeлeHйя дыя ймплatatHoгo 252Cf и peэлъtatл craвHиcв c имeюшимиcя экcПимeHtaлъHлми дahhыми. Пrиьoдяtcя taкжe кrивыe выcч;иtahhouBи изoдoзы Heйtrohoв ддя лиHeйhыX иctoч;hикoв длиhoй 1,35 и 2cм.

NEUTRON AND GAMMA-RAY DOSIMETRY IN AN ANIMAL EXPOSURE VOLUME AT A PULSED TRIGA REACTOR.

The distribution of steady-state and pulsed neutron and gamma-ray dose was measured in an animal exposure volume at the TRIGA reactor at the General Atomic facility. The usable volume was restricted to the lower 20 in. of an aluminum irradiation tube having an i.d. of 9 1/2 in. The tube was inserted into the pool adjacent to the reactor void tank, 134 cm from the core. Dosimetry films were used to determine gamma-ray dose while Sievert ionization chambers were used to determine neutron dose. Gold foils were used to measure the thermal neutron flux. Neutron doses were referenced to sulfur foil monitors. The neutron spectrum was examined with a system of threshold detectors and compared with spectra from Godiva II and the biological port of the Argonne CP-5 reactor. The vertical variation of dose along a lucite board, designed for retaining mice, was measured at the midplane of a lucite canister lowered into the tube. The dose in the lower 16 in. was found to vary less than 5 per cent. The ratio of neutron to gamma dose was approximately 6 to 1. A tissue-equivalent dog phantom, 20 in. long and 72 in. in diameter, was designed to determine the absorbed dose in depth. The fast neutron dose was about equal to the gamma-ray dose at the midline of the phantom having dropped from 2.3 times the gamma-ray dose in front of the phantom. The neutron plus gamma-ray dose at the midline was 44 per cent of the dose at the front surface of the phantom. Bilateral exposure doses were computed from the unilateral exposures.