Neutron Dose Rates Research Articles

Determination of dose components in phantoms irradiated with an epithermal neutron beam for boron neutron capture therapy

The application of activation foils, thermoluminescent detectors, and ionization chambers has been investigated for the determination of the different dose components in phantoms irradiated with a mixed gamma-ray and epithermal neutron beam for boron neutron capture therapy. The thermal neutron fluence has been determined using a set of AuAl and MnNi activation foils. TLD-700 and a Mg(Ar) ionization chamber have been used for the determination of the gamma-ray dose. The dose from epithermal neutrons has been determined using a TE(TE) ionization chamber. The detector characteristics and the relative sensitivities of the various detectors to the different dose components in the phantom have been determined. The following accuracies (1 standard deviation) in the determination of the different components have been obtained: thermal neutron fluence rate: 5%; gamma-ray dose rate: 7%; epithermal neutron dose rate: 15%. These values make these detectors suitable for obtaining the complete set of clinical dosimetry data required for patient dose assessment.

Gas Bremsstrahlung and Associated Photon-Neutron Shielding Calculations for Electron Storage Rings

The EGS4 electron-photon Monte Carlo code has been used to study the characteristics of the bremsstrahlung x rays generated from the interaction of circulating electrons with the residual gas in accelerator storage rings. Gas bremsstrahlung dose rates are given for various opening angles as a function of the electron beam energy ranging from 0.5-10 GeV. Photon and neutron dose rates, generated from various devices struck by gas bremsstrahlung in a synchrotron radiation beamline, are also presented along with the photon spectral and transmission results. The EGS4-predicted results are found to be in basic agreement with the measurements made at the Stanford Synchrotron Radiation Laboratory. Figures, equations, and a simple method useful for the photon-neutron shielding design for beamlines are provided.

Neutron generator at Hiroshima University for use in radiobiology study.

A neutron generator (HIRRAC) for use in radiobiology study has been constructed at the Research Institute for Radiation Biology and Medicine, Hiroshima University (RIRBM). Monoenergetic neutrons of which energy is less than 1.3 MeV are generated by the 7Li(p,n)7 Be reaction at proton energies up to 3 MeV. The protons are accelerated by a Schenkel-type-accelerator and are bombared onto the 7Li-target. An apparatus for the irradiation of biological material such as mice, cultured cells and so on, was designed and will be manufactured. Neutron and gamma-ray dose rates were measured by paired (TE-TE and C-CO2) ionization chambers. Contamination of the gamma ray was less than about 6% when using 10-microns-thick 7Li as a target. Maximum dose rates for the tissue equivalent materials was 40 cGy/min at a distance of 10 cm from the target. Energy distributions of the obtained neutrons have been measured by a 3He-gas proportional counter. The monoenergetic neutrons within an energy region from 0.1 to 1.3 MeV produced by thin 7Li or 7LiF targets had a small energy spread of about 50 keV (1 sigma width of gaussian). The energy spread of neutrons was about 10% or less at an incident proton energy of 2.3 MeV. We found that HIRRAC produces small energy spread neutrons and at sufficient dose rates for use in radiobiology studies.

Calculation of neutron dose rates including backscattering of a POLLUX cask in drifts of a repository mine

The Active Handling Experiment with Neutron Sources (AHE) is intended to study the effect of neutron backscattering from a salt environment during handling of highly active material (spent fuel and high level waste) in a repository located in a salt dome. The AHE project is planned to provide the calculated dose rates of a POLLUX cask with and without salt environment by comparison with the measured dose rates at a smaller experimental shielding cask loaded with Cf-252 neutron sources.

Calculated neutron and gamma dose rates around a moderated Am-Be neutron source

Neutron sources are widely used in Egypt for research and applications. All α-neutron sources produce fast neutrons and high energy γ-rays. In the present study, moderated Am-Be neutrons are examined and hence mixed field γ-rays and neutron dose equivalents are computed for various moderating water spheres. The results here presented indicate that the γ/ n dose equivalent ratio increases as the radius of the moderating sphere is increased. Hence, attention should be paid to γ-ray dose equivalent for radiation protection purposes.

The Biological Effectiveness of Neutrons: Implications for Radiation Protection

The radiobiology of neutrons has important societal implications. In nuclear facilities worldwide, several hundred thousand individuals are monitored as potentially receiving doses of neutrons, while there is increasing concern about the dose to which airline crew members are exposed, which includes a substantial neutron component. This represents an even larger number of individuals. The relative biological effectiveness (RBE) for neutrons varies with neutron energy. With cell killing as an endpoint, the maximum RBE occurs at about 300 to 400 keV, and falls off above and below this energy. For oncogenic transformation, the variation of effectiveness with energy is less marked. However, for some neutron energies and at some doses, there is a clear inverse dose-rate effect i.e. an increased rate of transformation with dose protraction, the magnitude of which depends on neutron energy, dose and dose rate. There is considerable uncertainty at the present time concerning the RBE of low energy 'soft' neutrons in the range 10 to 100 keV.

Design calculations for a local radiation shield of a radioisotope production target bombardment station

Multigroup discrete ordinates radiation transport calculations were performed to evaluate the present configuration of the local radiation shield of the target bombardment station for radioisotope production at the National Accelerator Centre. Calculated neutron and γ-ray dose rates and dose attenuation factors for this shield were compared with experimental measurements. Finally, a parametrical shield optimization study was performed.

Induction of Phenotypically Altered Mammary Epithelial Cells by Neutrons and γ Rays

These studies have examined alterations in the in vivo growth properties of mammary epithelial cells isolated at 1, 4, and 16 weeks after in vivo irradiation with -137Cs gamma rays or fission-spectrum neutrons. Altered in vitro growth potential was characterized by the proliferation of epithelial foci (EF) from irradiated animals under conditions in which mammary cells from nonexposed animals senesced. These EF were further characterized based on their ability to be subcultured. Both gamma and neutron irradiation resulted in the appearance of cells capable of forming EF. Further, with increased time in situ between irradiation and cell isolation, the frequency of EF which were capable of being subcultured indefinitely (EFs) increased. Reducing the gamma-ray dose rate resulted in fewer EFs while reducing the neutron dose rate resulted in increased frequencies of EFs. These data confirm earlier observations following gamma irradiation and show these cellular changes are also observed following neutron irradiation. In addition, these data indicate that changes in dose rate primarily influence the emergence of immortalized cell populations.

Physical, Biophysical, and Cell-Biological Factors That Can Contribute to Enhanced Neoplastic Transformation by Fission-Spectrum Neutrons

In radiobiology, fission-spectrum neutrons frequently have been used as a surrogate for other high-LET radiations, particularly when thick absorbers were involved as in animal studies. However, the spectrum of proton secondaries, plus the gamma rays generated in the absorption processes, suggests that a characterization of such a beam, based upon an average LET alone, may not adequately account for the spectrum of biological properties that it may have. Conflicting results have been reported on the relative effectiveness of reduced dose rates of fission-spectrum neutrons, and other high-LET radiations, for the induction of noeplastic transformation of cells in culture. Enhanced rates of neoplastic transformation were reported for C3H 10T1/2 mouse cells, Syrian hamster embryo cells, and human hybrid cells-all with the same beam of fission-spectrum neutrons generated by the JANUS reactor at the Argonne National Laboratory. No enhancement was observed with C3H 10T1/2 cells exposed to the beam from the TRIGA reactor at the Armed Forces Radiobiological Research Institute, or to maximally effective alpha particles. The recent report that an enhancement was also observed when human hybrid cells were exposed at a low dose rate to the TRIGA beam indicated that physical factors alone were not responsible for the differences observed with C3H 10T1/2 cells exposed to these various beams. To resolve the lack of consistency in the results that had been reported, a biophysical model was developed based, in part, on the existence of a narrow age interval in the growth cycle of a cell during which it is particularly sensitive to radiation neoplastic transformation. Because of the special physical and biological properties of cells in M phase, and/or in late G2 phase or early G1 phase, these cohorts of cells were proposed as those that are hypersensitive to neoplastic transformation by radiation.

Radiation Protection Metrology at a High-Energy Neutron Therapy Facility

A radiation protection survey has been carried out at a high-energy neutron therapy facility using a combination of different detectors and counters. Included in the survey were measurements with a tissue equivalent proportional counter (TEPC), a rem meter, a large volume ionisation chamber (LVI) and a Geiger counter. Dose equivalent rates, normalised to a proton beam current of 25 µA, between 1 µSv.h-1 and 0.7 Sv.h-1 were recorded depending on the location. In general the results confirm the tendency of the rem meter to over-read in fields consisting mainly of low energy neutrons and illustrate the advantages of the diagnostic and gamma discriminating properties of the TEPC. The LVI-Geiger system was found to be the least favourable combination of dosemeters, substantially under-reading and being unable to estimate the neutron dose rate at levels below about 32 µGy.h-1.

A Long-Term Study of Personal Neutron Monitors in Stray Fields Around High Energy Proton Accelerators

The results are presented of a long-term study of personal neutron monitors intended to be used in stray fields around high energy proton accelerators. The aim of this study was to investigate the possibility of using the NTA emulsion, CR-39, and Makrofol (polycarbonate) as personal monitors over extended periods given the low neutron dose rates outside the biological shielding of the accelerators operated at CERN. Considering the low doses recorded, a conservative interpretation of the detector result in terms of dose equivalent is nearly always possible, thus overcoming the problems of energy response of the detectors under test. The results show that any of the track detectors can be used for the purpose of long-term monitoring, so that the final decision on the choice is based on the costs involved.

Application of a germanium detector as a low flux neutron monitor

Abstract A feasibility study of using a germanium detector as a neutron monitor is described. A p-type hyperpure germanium detector with 10% relative efficiency is irradiated by neutrons with various energies. The p-type semiconducting detector emits prompt photopeaks at 596 and 691 keV from the Ge(n, γ) and Ge(n, n′γ) reactions, respectively. The count rate of these index photopeaks is used to monitor the neutron dose rate equivalent, resulting in a detection range of 0.08–120 μSv/h for thermal and epithermal neutrons and 0.8–75 μSv/h for fast neutrons. Advantages and disadvantages of using the HPGe detector as a neutron monitor while serving primarily as a gamma ray spectrometer in a mixed neutron/gamma field are discussed.

Application of FELTRAN to NEACRP TN12 shipping cask benchmark

The finite element technique has received attention as a method of solving the Boltzmann equation in one, two and three spatial dimensions. In particular much work has been published concerning the geometrical flexibility, speed and accuracy of the method. British Nuclear Fuels plc and Imperial College have collaborated in developing the finite element code FELTRAN to near production code status. FELTRAN solves the even parity form of the Boltzmann equation using a functional approach. The solution is found in one or two spatial dimensions using various orders of finite elements to specify the problem geometry. The angular dependence of the even parity flux is expressed using spherical harmonics. FELTRAN has been interfaced to ANISN formatted nuclear data libraries such as CASK and BUGLE. Anisotropic scattering may be specified to any order. Methods have been incorporated within the code to analyse systems with voids. FELTRAN is currently undergoing further development as part of a BNFL sponsored MSc research programme at the University of Salford. The purpose of this paper is to consider the application of FELTRAN to a practical shield design problem. The OECD have adopted a benchmark experiment to measure the neutron and gamma ray radiation dose rates around a spent fuel transport flask. As part of an international collaboration the physical details of the flask design and contents have been provided to the nuclear industry. The objective is to perform an international comparison of the methods used in the analysis of cask shielding. BNFL is one of the companies involved, using the well established codes RANKERN and MCBEND. The FELTRAN calculations are performed using the same source and geometry data and equivalent angular flux expansions as for these two codes. FELTRAN is then compared with experimental and calculated results.

Neutron measurements around medical electron accelerators by active and passive detection techniques

Passive and active detection techniques have been employed in order to measure neutron fluence rates and corresponding exposure rates around medical electron accelerators operating at energies well above the neutron binding energies of the structural materials. In these conditions from the treatment head, in the direct photon flux and from the shielded region, a fast neutron flux emerges which is partly absorbed and partly scattered by the walls, eventually establishing a nearly uniform thermal and epithermal flux in the room. Both direct and scattered flux contribute to the dose to the patient. A smaller neutron dose rate can also be found outside the treatment room, where the therapy staff works. Passive detectors, of moderation type, have been employed in the treatment room and 3He active detectors in the external zones. For the treatment room the activation data were compared with results of Monte Carlo simulation of the neutron transport in the room. Technical features of the two measures are briefly presented and results obtained around three different types of accelerators are reported. At the higher beam energies, i.e., 25 MV, a neutron dose of 0.36 Sv was estimated in the treatment field in addition to a therapeutic x-ray dose of 50 Gy. At lower energies or out of the treatment field the neutron dose drops significantly. In the external zones the dose rates everywhere are below the recommended limits and normally very low, the highest values being recorded in positions very close to the access door of the treatment room.

Theoretical neutron flux levels, dose rates, and metal foil activation in electrochemical cold fusion experiments

Theoretical calculations of neutron flux rates and the corresponding neutron dose rates were performed for various surfaces of our electrochemical calorimetric cell. This cell consisted of a palladium rod cathode surrounded by a platinum coil anode in a glass tube that initially contained 18-20 g of 0.1 M LiOD+D 2 O

Neutron and gamma ray measurements carried out around a Varian Clinac 2500 linear accelerator

Auckland Hospital, New Zealand, has recently installed a Varian Clinac2500 linear accelerator. This unit is capable of producing 24 MV peak photon energies. It is well known (Fox et al, 1977; Ing et al, 1982; Palta et al, 1984) that such photons are capable of producing neutrons via a photonuclear reaction. The sources of such neutrons from this accelerator have been described (LaRiviere, 1985). In this study a survey was carried out around this new installation to observe some of the characteristics of the neutrons produced, and to assess likely radiological hazards to radiotherapy staff. Equipment and methods Equivalent neutron dose was measured using an Alnor neutron dose rate meter (rem-meter) model 2202D. Remote pulse counting was carried out using an Elscint model INS-30 nuclear spectrometer. The manufacturer of this unit claims that the detector exhibits 200 R/h gamma ray discrimination. Initially this would appear satisfactory for carrying out a survey, but owing to the fact that linear accelerators are operatedin pulsed mode, peak leakage and scatter photon dose rates within the treatment room could well exceed 200 R/h. This was confirmed by operating the accelerator in 6 MV photon mode (i.e. producing no expected neutrons) and showing that the neutron meter did in fact respond. However, in the entrance maze to the treatment room the scattered X-ray dose rate was attenuated sufficiently so that it was not contributing to detector response.

Application of new calculation methods to the primary shielding of the avr experimental reactor

The DOT/RATRAC code system has been verified by comparison with gamma and neutron dose rates and neutron energy spectra measured at AVR experimental reactor (HTGR-type reactor). The RATRAC module was recently developed for improving transport calculations for complex configurations containing extended voids and for saving computing time. In RATRAC the angular flux density at the surfaces of the voids is described in terms of spherical harmonics and the radiation transport through the voids is treated analytically. It turns out, that the theoretical results and the values measured at the AVR reactor are in good agreement, even on top of the reactor building.

Application of the Monte Carlo Method for Neutron Shielding Analysis of Transport Casks

AbstractThe Monte Carlo Method using the next-event surface crossing (NESX) estimator and the Monte Carlo coupling technique were applied to anillyse the neutron shielding problems of transport casks and the neutron dose rate distribution in a spent fuel shipping vessel. The subject of the experiment, the shielding of a spent fuel shipping cask, was carried out at the Central Research Institute of Electric Power Industry (CRIEPI), Japan, on-board a spent fuel shipping vessel, the Pacific Swan, in which 13 casks of TN-12A and Excellox-3 were loaded. The optimum shielding arrangement for a cask shielding system was also studied. With the present knowledge of Monte Carlo techniques the method could be employed as an option for analysing the radiation shielding of transport casks. In addition, the present experimental data can be adopted as benchmarks for transport casks and vessel design.

Optimization of boron and neutron delivery for neutron capture therapy.

A number of groups in the United States have received funding that will permit evaluation of the clinical efficacy of the neutron capture therapy (NCT) procedure. Various reactors are being modified to allow the construction of an epithermal neutron beam. At the Brookhaven Medical Research Reactor (BMRR), the patient irradiation facility is being modified to produce an optimized epithermal neutron beam. An 80-cm-thick A1-D2O mixture (184 g/cm2, 25% D2O by volume) is being installed in the shutter assembly. One-dimensional calculations indicate that this configuration should provide an epithermal neutron flux density of approximately 1 x 10(9) n/cm2/sec at 3 MW and a concomitant fast neutron dose rate of approximately 2 x 10(-11) rad per epithermal neutron (assuming a homogeneous A1-D2O mixture). The actual geometry will be an inhomogeneous array of D2O and A1 layers producing parameters somewhat less favorable than those listed above; experimental verification is in progress. Significant gains have recently been made in selectively targeting B to melanoma with various melanaffinic compounds, including p-boronophenylalanine, and with boronated porphyrins that may be applicable to a variety of tumors. Neutron capture radiographs have been obtained with the above compounds, and efforts have been made to quantitate boron uptake in growing and quiescent or necrotic regions of tumor via double-labeling techniques obtained with tritiated thymidine. A correlation between therapeutic efficacy and the ability to deliver boron to viable areas of tumor has been observed.

Optimum Arrangement to Minimize Total Dose Rate of Iron-Polyethylene Shielding System

The shielding experiments using iron-polyethylene slab shields with a 252Cf neutron source were carried out to find out an optimum arrangement to minimize the total dose rate; that is composed of neutron and secondary γ-ray dose rates. The total thickness of the iron slabs was fixed at 32 cm, while a variety of thickness and location of a polyethylene slab in the iron slabs were employed as a parameter. The minimum dose point (i.e. optimum shielding arrangement) was observed when the polyethylene slab was located at approximately 20 cm depth from the source side in the arrangements. The ratio of the minimum dose rate obtained for the optimum arrangement to the maximum for the worst arrangement became 1/1.3 for the polyethylene slab of 1-cm-thick, 1/2.0 for 3-cm-thick, 1/2.9 for 6-cm-thick, 1/3.9 for 10-cm-thick and 1/3.6 for 14-cm-thick. The appearance of the optimum arrangement for the total dose rates, and changing profiles of the secondary γ-ray as well as the neutron dose rates were reproduced for the...