Development of a neutron detector for radiation protection monitoring

The paper presents the results of the development of a neutron detector for radiation protection purposes. Monte Carlo simulations, using MCNP5 code, were performed to optimize the configuration of the neutron detector. The developed detector consists of a 3He proportional counter embedded in a multi-layer moderator made of high-density polyethylene (HDPE) and Cadmium. The characteristics of the developed neutron detector including neutron fluence response and ambient dose equivalent response were calculated, analyzed and compared with those from other neutron survey meters. The simulation model and computed results were assessed through experimental measurements at the Secondary Standards Dosimetry Laboratory of the Institute for Nuclear Science and Technology (INST). A good agreement between the simulated and experimental results was observed within 9.3% for 241Am-Be source and four simulated workplace neutron fields.

Simulation of absorbed doses distribution in a polyethylene phantom for BNCT application at the Dalat research reactor

In this paper, the distribution of absorbed dose components in a polyethylene phantom for BNCT application at Dalat Nuclear Research Reactor (DNRR) were calculated using the MCNP5 code. The configuration of horizontal neutron channel No.2 of the DNRR, which contains a cylindrical collimator with neutron filters of 20-cm Si and 3-cm Bi, was simulated. The results show that the gamma dose along the central axis of the phantom has the maximum value of 1.82×10-6 Gy at the 0.5-cm depth, and reduces to 9.05×10-7 Gy at the 3-cm depth. The main contribution to gamma dose is due to the interaction of thermal neutron with hydrogen in the phantom via the 1H(n,γ)2H reaction, and its value is much smaller than thermal neutron dose. The total absorbed dose along the central axis of the phantom has the maximum value of 7.87×10-5 Gy at the 0.5-cm depth, and decreases rapidly to 1.52×10-5 Gy at the 3-cm position, and mainly depends on the boron and thermal neutron doses caused by the 10B(n, α)7Li and 14N(n, p)14C reactions, respectively.

Fullerite Nano-materials as a Moderator in Neutron Irradiators with Radium Sources: Feasibility and Advantages

In this study, the feasibility of using fullerite nano-materials as a moderator in 226Ra-Be neutron irradiators has been theoretically investigated, for the first time. Thermal, intermediate and rapid neutron flux in irradiation channels was calculated using the MCNP5 code when a fullerite nano-material was used as a moderator. The simulation results were then compared with other simulation results performed when paraffin was used as a moderator. The comparison showed that using fullerite instead of paraffin as a moderator results in an increase in the total number of irradiation neutrons by more than twice in average (240 %) for each direction inside the irradiator. This increase is distributed as follows: 27.84 %, 87.84 % and 124.32 % thermal, intermediate and rapid neutrons, respectively. The previous distribution indicates a significant increase in the intermediate and fast neutron flux. This is considered as an additional advantage of using the 226Ra-Be neutron irradiator with a fullerite moderator. The irradiator can then be used not only to irradiate the materials whose irradiation requires thermal neutrons, but also those that require medium- to high-energy neutrons. Keywords: Neutronic irradiator, Ra-Be radiation source, Cadmium, Neutron flux, MCNP5-beta code. PACS: Neutrons diffusion and moderation, 28.20.Gd, Moderators (nuclear reactors), 28.41.Pa.

Angular Distribution of Neutrons Around Thick Beryllium Target of Accelerator-Based 9Be(d, n) Neutron Source

An experimental work and simulations were carried out to determine the angular distributions of neutrons and yields of the 9Be(d, n) reaction over all angular range (360 deg) on a thick beryllium target as an accelerator-based neutron source at incident-deuteron energy 13.6 MeV. The neutron activation method was used in the experimental part using aluminum and iron foils as detectors to calculate the neutron flux. The Monte Carlo neutral-particles code (MCNP5) was used to demonstrate and simulate the neutron distribution, also to understand and compare with the experimental results. The neutron energy spectrum was computed using the projection angular-momentum coupled evaporation code PACE4 (LISE++) and the spectrum was adopted in MCNP5 code. Two experimental ways were used, one with beryllium target and another one without the beryllium target, to evaluate the neutron flux emitted only by the beryllium target. Typical computational results were presented and are compared with the previous experimental data to evaluate the computing model as well as the characteristics of emitted neutrons produced by the 9Be(d, n) reaction with a thick Be-target. Moreover, the results can be used to optimize the shielding and collimating system for neutron therapy.

Modeling an HPGe detector response to gamma-rays using MCNP5 code

The response function is the important information for the precise interpretation of experimental data and also for characterizing the developing nuclear instruments. Measurement of the response function normally requires a number of mono-energetic gamma-ray sources, a long acquisition time and an appropriate experimental setup. The Monte Carlo method, as an alternative to response function measurement, has widely been used and recommended. In this study, a computational model of an HPGe detector has been developed by using the MCNP5 code. To validate the simulated model, the simulations from mono-energetic sources have been compared to the corresponding measured data. Any deviation from the measurement could be attributed to the unmodeled details of the detector crystal, so they needed adjustment. Moreover, an analysis has been undertaken on the dependency of detection efficiency on the dead layer thickness of the germanium crystal. Having developed a computational model of the crystal, a set of correction factors was extracted to take into account the gamma-ray self-absorption within the source volume. The simulated model of the HPGe detector in this study can be used to calculate the detection efficiency when the samples are not of the standard geometry which require self-absorption considerations.

197Au(n,xn) reactions at The Svedberg Laboratory high-energy neutron facility in Uppsal

Cross section measurements for the197Au(n,xn) reactions have been performed at The Svedberg Laboratory (TSL) high-energy neutron facility in Uppsala,Sweden. The 45.6 and 58.3 MeVquasi-monoenergetic neutron beams were produced by means of the 7Li(p,n) reaction and were monitored with thin-film breakdown counters (TFBCs). After the end of the irradiations, the activity induced by the neutron beams in the targets and in reference foils, has been measured by a HPGe detector. In order to determine the cross sections of the (n,xn) reactions, the spectral neutron flux distribution is needed, thus the characterization of the beam is of major importance. Therefore, simulations that take into account the whole experimental setup of the irradiation have been performed with the use of MCNP5 code and the results are presented in this work. Currently, further analysis of the data is in progress.