Recent developments of plastic scintillators are re-viewed, from 2000 to March 2014, distributed in two different chapters. First chapter deals with the chemical modifications of the polymer backbone, whereas modifications of the fluorescent probe are presented in the second chapter. All examples are provided with the scope of detection of various radiation particles. The main characteristics of these newly created scintillators and their detection properties are given.

Current Status on Plastic Scintillators Modifications

We would like to present here with the eyes of the chemist the most recent developments of plastic scintillators (PS) for neutron detection. This review covers the period from 2000 to August 2014, and is fragmented in two main chapters. The first chapter deals with the chemical modifications for thermal neutron capture, whereas the second chapter presents the various strategies used to enhance the response to fast neutrons via pulse shape discrimination. For each chapter the theory is also explained.

Pulse shape discrimination between (fast or thermal) neutrons and gamma rays with plastic scintillators: State of the art

Natural gadolinium displays a number of remarkable physical properties: it is a rare earth element, composed of seven stable or quasi-stable isotopes, with an exceptionally high magnetization and a Curie point near room temperature. Its use in the field of nuclear instrumentation historically relates to its efficiency as a neutron poison in power reactors. Gadolinium is indeed the naturally occurring element with the highest interaction probability with neutrons at thermal energy, shared between Gd-157 (15.65%, 254000 b cross section) and Gd-155 (14.8%, 60900 b) isotopes. Considering that neutron capture results in an isotopic change, followed by a radiative rearrangement of nuclear and atomic structures, Gd may be embodied not merely as a neutron poison but as a neutron converter into a prompt photon and an electron source term. Depending on the nature and energy of the reaction products (from a few-keV Auger electrons up to 8 MeV gamma rays) that the detector aims at isolating as an indirect neutron signature, a variety of sensor media and counting methods have been introduced during the last decades. This review first draws a theoretical description of the radiative cascade following Gd ( n , γ ) capture. The cascade may be subdivided into regions of interest, each corresponding to dedicated detection designs and optimizations whose current status is detailed. This inventory has allowed the authors to extract and benchmark key figures of merit for the definition of a detection scheme: neutron attenuation, neutron sensitivity (cps/nv), gamma rejection, neutron detection limit in a mixed field, intrinsic or extrinsic moderation, and transportability. On this basis, the authors have identified promising paths for Gd-based neutron detection in contemporary instrumentation.

Gadolinium for neutron detection in current nuclear instrumentation research: A review

The various instruments used for neutron spectrometry can be divided into four categories. Within each of these categories, the underlying measurement principle for all devices is essentially the same. The applications of the instruments within each group to spectrometry for radiation protection are reviewed.

Neutron spectrometry for radiation protection.

A basic PGNAA facility with an Am–Be neutron source is described to analyze the pollutants in water. The properties of neutron flux were determined by MCNP calculations. In order to determine the efficiency curve of a HPGe detector, the prompt-gamma rays from chlorine were used and an exponential curve was fitted. The detection limits for typical water sample are also estimated using the statistical fluctuations of the background level in the areas of recorded the prompt-gamma spectrum.

Detection limits of pollutants in water for PGNAA using Am-Be source

Accurate estimation of neutron dose requires knowledge of the neutron energy distribution in the working environment. Existing neutron spectrometry systems, Bonner spheres for example, are large and bulky, and require long data acquisition times. A portable system that could indicate the approximate neutron energy spectrum in a short time would be extremely useful in radiation protection. A composite scintillator, consisting of lithium gadolinium borate crystals in a plastic scintillator matrix, produced by Photogenics is being tested for this purpose. A prototype device based on this scintillator and digital pulse processing electronics has been calibrated using quasi-monoenergetic neutron fields at the low-scatter facility of the UK National Physical Laboratory (NPL). Energies selected were 144, 250, 565, 1400, 2500 and 5000 keV, with correction for scattered neutrons being made using the shadow cone technique. Measurements were also made in the NPL thermal neutron field. Pulse distributions collected with the digitiser in capture-gated mode are presented, and detection efficiency and energy resolution derived. For comparison, neutron spectra were also collected using the commercially available Microspec N-Probe from Bubble Technology Industries, which consists of an NE213 scintillator and a 3He proportional counter.

Response of a lithium gadolinium borate scintillator in monoenergetic neutron fields.

Lithium-gadolinium-borate (LGB) dispersed as microcrystals within the plastic scintillator BC-490 is a promising material for accurate neutron dosimetry in mixed n/gamma fields. Spectral information > 1 MeV is obtained by capture gating proton recoil events in the plastic scintillator to subsequent capture in (6)Li. Below 1 MeV, isolated capture events in either gadolinium or (6)Li give energy information in this region. Discrimination based on capture gating is used to reject false coincidences due to gamma rays or incorrectly gated neutron events. A detailed Monte Carlo model has been created in MCNPX that predicts the energy response of the LGB spectrometer in the capture-gated mode of operation. X-ray microtomography has been performed on the detector in order to obtain the LGB microcrystal distribution within the plastic scintillator, and this is incorporated into the model. The way in which the calculated response functions can be included in an unfolding procedure is outlined.

Lithium–gadolinium–borate as a neutron dosemeter

The Transportable Neutron Spectrometer (TNS) has been used by the Ministry of Defence for over 15 years to characterise neutron fields in workplace environments and provide local correction factors for both area and personal dosimeters. In light of advances in neutron spectrometry, a programme to evaluate and improve TNS has been initiated. This paper describes TNS, presents its operation in known radioisotope fields and in a reactor environment. Deficiencies in the operation of the instrument are highlighted, together with proposals for updating the response functions and spectrum unfolding methodologies.

Considerations in the design of an improved transportable neutron spectrometer

An activation analysis facility based on an isotopic neutron source (185 GBq 241Am/Be) which can perform both prompt and cyclic activation analysis on bulk samples, has been used for more than 20 years in many applications including 'in vivo' activation analysis and the determination of the composition of bio-environmental samples, such as, landfill waste and coal. Although the comparator method is often employed, because of the variety in shape, size and elemental composition of these bulk samples, it is often difficult and time consuming to construct appropriate comparator samples for reference. One of the obvious problems is the distribution and energy of the neutron flux in these bulk and comparator samples. In recent years, we have attempted to adopt the absolute method based on a monostandard and to make calculations using a Monte Carlo code (MCNP4C2) to explore this further. In particular, a model of the irradiation facility has been made using the MCNP4C2 code in order to investigate the factors contributing to the quantitative determination of the elemental concentrations through prompt gamma neutron activation analysis (PGNAA) and most importantly, to estimate how the neutron energy spectrum and neutron dose vary with penetration depth into the sample. This simulation is compared against the scattered and transmitted neutron energy spectra that are experimentally and empirically determined using a portable neutron spectrometry system.

Monte Carlo calculations and neutron spectrometry in quantitative prompt gamma neutron activation analysis (PGNAA) of bulk samples using an isotopic neutron source

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Papers

Response of a lithium gadolinium borate scintillator in monoenergetic neutron fields.

Lithium–gadolinium–borate as a neutron dosemeter

Considerations in the design of an improved transportable neutron spectrometer

Monte Carlo calculations and neutron spectrometry in quantitative prompt gamma neutron activation analysis (PGNAA) of bulk samples using an isotopic neutron source