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In this study, four alloys composed of highly pure elements (lead, zinc, tin and bismuth), named Pb–Zn50, Pb–Sn50, Sn–Zn50 and Sn–Bi50 were prepared by rapid solidification process (RSP) using melt-spun technique. The structural, mechanical properties (elastic and plastic deformations), and microhardness have been investigated and analyzed using X-ray diffraction, tensile test machine, and vickers microhardness respectively. According to results, the yield strength, ultimate tensile strength (UTS), toughness, and Vickers hardness (Hv) were improved due to the advantages of rapid solidification technology including elimination of grain boundary segregation, improvement in alloy homogeneity, refinement of grain size, formation of novel metastable crystalline, rises in the equilibrium solid solubility's of solute elements and reduction in the degree of order. This can be ascribed to reinforcement of Zn and Bi particles, refined β-Sn, Zn and Bi grains that could obstruct the dislocation slipping. In addition, the shielding performance of these alloys for gamma radiation were examined and analyzed utilizing WinXCom software over energy range 0.015–15 MeV. The alloys under study found to be good shielding materials against γ-ray relative to common shielding materials and newly studied materials. The results indicate that the Sn–Bi50 alloy has the highest values of μm, Z eff and Z eq ranging from 0.039 to 81.3 cm2/g, 60.41–73.83, and 64.02–70.793 respectively. While it has the smallest values of HVL (ranging from 8.73E-4 to 1.86 cm) which means that it is the best choice as a γ-ray shielding material. In aspects of neutron attenuation characteristics, the results also indicate that the Pb–Zn50 alloy has the highest value of Σ R (0.133 cm−1) amongst all other alloys, frequently used neutron shielding materials, and recently studied materials. Furthermore, proton (H+1) and alpha particles (He+2) projected range (PR) and mass stopping power (MSP) values have been calculated. It can be concluded that Zn–Pb alloy among the examined samples is superior in terms of shielding against proton and alpha particles. This paper provides a comprehensive examination of the use of lead and the development of new safety lead-free shielding materials in many applications including nuclear medicine. • Conventional and lead-free binary alloys. • Rapid solidification process using melt-spun technique. • Structural features, microhardness and mechanical properties. • X-ray diffraction (XRD), Vickers microhardness and tensile test machine techniques. • X/Gamma-rays, fast neutron, proton and alpha particles shielding features. [ABSTRACT FROM AUTHOR]

The present paper reports the thermoluminescence (TL) based personnel neutron dosimetry studies of LiMgBO 3 : Dy3+ phosphor. It offers the great advantage of being a near tissue equivalent material with Z eff = 9.1. Also, the presence of both Li and B in the host matrix directly induce TL signal for neutrons. This is highly advantageous over CaSO 4 : Dy based TL neutron dosimeters where it requires to mix some thermal neutron sensitive species with the host matrix to induce neutron response. LiMgBO 3 : Dy3+ was synthesized by sol gel method and its structural characterizations were done by XRD, SEM and EDS techniques. TL upon irradiation with neutrons showed the glow peak at ~140 °C (413K). TL response up to 200 mSv of neutron dose showed very good linearity with adj. R2 = 0.998. 30% borated rubber discs were successfully utilised for the segregation of the gamma dose from the total dose in the mixed field (neutron + gamma) environments. The minimum detectable dose of the material was found to be 27.5 μGy and 2.7 μSv for gamma and thermal neutrons respectively. The TL sensitivity of LiMgBO 3 : Dy3+ for neutron and gamma radiation is 0.6 times and 0.3 times respectively with respect to that of standard TLD-100 [LiF: Mg, Ti] material. The proposed LiMgBO 3 :Dy3+ was also found to have better n/γ dose discrimination capability as compared to the commercially available TLD-100 material [LiF: Mg, Ti]. The kinetic parameters (viz. E, b and s) of the neutron irradiated TL glow peak were determined by Chen's peak shape and by variable heating rate methods. [ABSTRACT FROM AUTHOR]

Abstract: DEXA (dual-energy X-ray absorption) is widely used in airport scanners, industrial scanners and bone densitometers. DEXA determines the properties of materials by measuring the absorption differences of X-rays from a bremsstrahlung tube source with and without filtering. Filtering creates a beam with a higher mean energy, which causes lower material absorption. The absorption difference between measurements (those with a filter subtracted from those without a filter) is a positive number that increases with the effective atomic number of the material. In this paper, the concept of using a filter to create a dual beam and an absorption difference in materials is applied to radiation from a 252Cf source, called NEUDEG (neutron and dual-energy gamma absorption). NEUDEG includes absorptions for fast neutrons as well as the dual photon beams and thus an incentive for developing the method is that, unlike DEXA, it is inherently sensitive to the hydrogen content of materials. In this paper, a model for the absorption difference and absorption sum in NEUDEG is presented using the combined gamma ray and fast neutron mass attenuation coefficients. Absorption differences can be either positive or negative in NEUDEG, increasing with increases in the effective atomic number and decreasing with increases in the hydrogen content. Sample sets of absorption difference curves are calculated for materials with typical gamma-ray and fast neutron mass attenuation coefficients. The model, which uses tabulated mass attenuated coefficients, agrees with experimental data for porcelain tiles and polyethylene sheets. The effects of “beam hardening” are also investigated. [Copyright &y& Elsevier]

Abstract: One of the most important questions to be answered with regard to digital pulse-shape discrimination (PSD) systems based on organic scintillators is: What sampling properties are required for a fast-waveform digitizer used for digitizing neutron/gamma-ray pulses, while an accurate PSD is desired? Answering this question is the main objective of this paper. Specifically, the paper describes the influence of the resolution and sampling frequency of a waveform digitizer on the PSD performance of organic scintillators. The results presented in this paper are meant to help the reader choosing a waveform digitizer with appropriate bit resolution and sampling frequency. The results presented here show that a 12-bit, 250-MHz digitizer is a good choice for applications that require good PSD performance. However, when more accurate PSD performance is the main requirement, this paper presents PSD figures of merit to qualify the impact of further increasing either sampling frequency or resolution of the digitizer. [Copyright &y& Elsevier]

Abstract: Gamma ray tracking is an efficient detection technique in studying exotic nuclei which lies far from beta stability line. To achieve very powerful and extraordinary resolution ability, new detectors based on gamma ray tracking are currently being developed. To reach this achievement, the neutron–gamma discrimination in these detectors is also an important task. In this paper, by suitable layered feedforward neural networks (LFNNs), we have constructed novel and consistent empirical physical formulas (EPFs) for some highly nonlinear detector counts measured in neutron–gamma discrimination. The detector counts data used in the discrimination was actually borrowed from our previous paper. The counts used here had been originally measured versus the following parameters: energy deposited in the first interaction points, difference in the incoming direction of initial gamma rays, and finally figure of merit values of the clusters determined by tracking. The LFNN–EPFs are of explicit mathematical functional form. Therefore, by various suitable operations of mathematical analysis, these LFNN–EPFs can be used to derivate further physical functions which might be potentially relevant to neutron–gamma discrimination performance of gamma ray tracking. [Copyright &y& Elsevier]

• A technique for the clustering of electrical signals is proposed. • The clustering algorithm has been customized for the electrical pulses from high-energy particle detectors. • Clustering has been applied on the electrical signal coming from scintillator under radiation. • The ability to identify neutrons and gamma patterns has been verified. In several nuclear applications, scintillators, coupled with a photomultiplier and pulse amplifier, are used in order to detect high energy particles, i.e. neutrons and gamma rays. The different particles incident on the scintillator produce electrical pulses having different shape; moreover, the amplitude of these signals is related to the particles energy. The electrical pulses of the scintillator chain are acquired by digital systems that, generally, perform a triggered acquisition consisting of a stream of pulse windows. The aim of this study is the development of a simplified clustering algorithm able to produce reference patterns in compliance with the pattern recognition algorithm based on the matched filter technique, starting from a stream of pulses generated by particles having different energy and type. This paper contains a general description of the clustering algorithm and of the main customizations performed for the scintillator signals. In order to test in real case the efficiency, the algorithm has been applied on the data acquired during a radiation test performed at Frascati Neutron Generator for Stilbene scintillator. The results show that this algorithm works properly, deriving the centroids of the clusters representing the neutron and gamma shapes, together with their occurrences in the analysed data stream. [ABSTRACT FROM AUTHOR]

The Prompt Gamma Neutron Activation Analysis (PGAA) facility will be commissioned at the NB1 beam tube of the Moroccan TRIGA MARK II research reactor. PGAA is a non-destructive analytical technique used for multi-elemental analysis and especially for light element analysis. For better performances, PGAA requires small and focused beam, and high Ф th /Ф ep and Ф th /Ф fast ratios, and low background gamma radiation. For this purpose, fast neutron filters are commonly used to reduce fast and epithermal components of neutron flux. In this paper, the impact of sapphire filter, introduced, within the collimator hole, at the beam port on neutron spectra was evaluated. Monte Carlo, MCNP6.2, code has been used with ENDF VII.1 neutron cross section library to perform this study. An approximate TRIGA MARK II reactor MCNP model has been used to generate neutron spectra at the entrance of the collimator. Then an independent model for both collimator and sapphire filter has been brought into calculation to fulfil our purpose. Results showed an increase of about 8 times for both Ф th /Ф ep and Ф th /Ф fast ratios using a 10 cm length of sapphire. • Impact Sapphir (Al2O3) filter in components of the neutron beam. • Design of the collimator for PGAA. • Simulation of the collimator with Monte Carlo N-Particle code MCNP6.2 • Generate thermal neutron treatment file S (α , β) for Sapphir (Al2O3) using LEAPR/NJOY2016 [ABSTRACT FROM AUTHOR]

Epoxy resin is cheap, flexible, durable and abundant material. Within this framework, it is a very useful option in a comprehensive application field. In this paper, neutron shields based on epoxy based composite materials were studied by adding different percentages of additive materials as Fe (Iron), Bi (Bismuth), Ta (Tantalum) and WC (Tungsten Carbide). GEANT4 and FLUKA simulation codes were applied simulating and comparing the interactions of 4.5 MeV 241Am-Be neutrons and 1.25 MeV 60Co source energetic gamma rays with the different specific materials. The best shielding performance results were obtained with the Tantalum-Epoxy sample but all the studied samples demonstrate a good quality as neutron and gamma shielding, when compared to previous results found in literature. Implications of neutron and gamma bombardment were also discussed in the view of a large usage fields. • Improving flexible metal powder filled epoxy composite radiation shields. • Epoxy/ Fe, Bi, Ta, WC composites were designed by GEANT4 and FLUKA codes. • Simulations were applied for 4.5 MeV neutrons and 1.25 MeV gamma rays. • Epoxy/Ta composite has high shielding performance for neutrons and gamma rays. • Composites are high quality neutron and gamma shields as compared to the literature. [ABSTRACT FROM AUTHOR]

Highlights • Investigation of the influence of homogenization on nuclear heating and shielding. • Heterogeneous model with detailed description of the BSM cooling system. • Homogeneous model simplified by removing pipes and introducing mixed materials. • Nuclear heating was calculated for several components and on a fine mesh for both models. • Homogenization tends to underestimate the local heat deposition and overestimates shielding. Abstract The electron cyclotron-heating upper launcher (ECHUL) will be installed in four upper ports of the ITER tokamak thermonuclear fusion reactor. Each ECHUL is able to deposit 8 MW power into the plasma for plasma mode stabilization via microwave beam lines. These beam lines and other components needed for the guidance of the microwaves are mounted into the upper port plug. In order to protect these components and to mitigate radiation leaving the reactor, several shield blocks are also included in the port plugs design. The ECHUL components which are closest to the fusion plasma are heated by the gamma and neutron radiation and therefore will need to be actively cooled. This paper reports on the influence of material homogenization by comparing results obtained in the neutronic analyses performed for the cooling system of the ECHUL blanket shield module (BSM). A small reduction in shielding performance of the BSM was found in the case of a heterogeneous description of the BSM compared to the standard approach of homogenizing the materials of the BSM by removing pipes and cooling circuits for the neutronic analyses. [ABSTRACT FROM AUTHOR]

Abstract: Strong fields of thermal and epithermal neutrons as used for Boron Neutron Capture Therapy (BNCT) show a relatively small gamma dose. However, for accurate dose estimation, both the neutron as well as the gamma component has to be considered. Different approaches have been studied with various thermoluminescence detectors (TLDs). As a result, CaF2:Tm detectors (TLD-300) are attributed to be most insensitive to thermal neutrons but not tissue-equivalent. In this paper we focus on the measurement of the gamma dose with TLD-300 in mixed irradiation fields. To overcome the disadvantage of being not tissue-equivalent, a study on energy compensation filters was performed. The paper comprises also calibrations of TLD-300 using energy compensation filters at the PTB-reference field for thermal neutrons, and first test experiments at the TRIGA Mainz including simulations. [Copyright &y& Elsevier]

In this paper, different amounts of Szaibelyite (100, 300, 500, 700 and 900) were used to prepare Szaibelyite/epoxy resin composites for neutron shielding. Microstructures of fracture surfaces of the prepared composites were tested by scanning electron microscope while the neutron shielding properties were determined by Am-Be neutron source. It was found that the shielding percentage of all composites increases with the addition of Szaibelyite as well as the thickness of composites increment. M5 was found to be the best neutron shielding composite, the shielding percentage was nearly 100% when the thickness of M5 chosen to be 1.5 cm. Furthermore, the shielding mechanism of M5 for neutron shielding could be concluded as follows: Firstly, H, O, C, Fe and B-11 isotope contained in M5 slowed down the neutron to thermal neutron, which then absorbed by the B-10 isotope. According to the achieved results, the performance of the new neutron shielding material owns the value of practical application. [ABSTRACT FROM AUTHOR]

Time of flight experiments are the gold standard method for generating labeled training and testing data for the neutron/gamma pulse shape discrimination problem. As the popularity of supervised classification methods increases in this field, there will also be increasing reliance on time of flight data for algorithm development and evaluation. However, time of flight experiments are subject to various sources of contamination that lead to neutron and gamma pulses being mislabeled. Such labeling errors have a detrimental effect on classification algorithm training and testing, and should therefore be minimized. This paper presents a method for identifying minimally contaminated data sets from time of flight experiments and estimating the residual contamination rate. This method leverages statistical models describing neutron and gamma travel time distributions and is easily implemented using existing statistical software. The method produces a set of optimal intervals that balance the trade-off between interval size and nuisance particle contamination, and its use is demonstrated on a time of flight data set for Cf-252. The particular properties of the optimal intervals for the demonstration data are explored in detail. [ABSTRACT FROM AUTHOR]

Gamma-ray emission from special nuclear material (SNM) is relatively easy to shield from detection using modest amounts of high-Z material. In contrast, fast-neutrons are much more penetrating and can escape relatively thick high-Z shielding without losing significant energy. Furthermore, fast neutrons provide a clear and unambiguous signature of the presence of SNM with few competing natural background sources. The challenge of detecting fast neutrons is twofold. First, the neutron flux from SNM are only a fraction of the corresponding gamma-ray flux. Second, fast neutrons can be difficult to differentiate from gamma rays. The ability to discriminate gamma rays from neutrons, combined with a neutron imaging capability, can in some situations enable the localization of SNM neutron sources distinct from background. With the developments of pulse-shape-sensitive plastic scintillators that offer excellent gamma-ray/neutron discrimination, and arrays of silicon photomultipliers combined with highly scalable and fast positron-emission-tomography (PET) multi-channel readout systems, field-deployable neutron imagers suitable for SNM detection might now be within reach. In this paper, we present a characterization of the performance of a recently available commercial PET-scanner readout, including its sensitivity to pulse-shape differences between fast neutrons and gamma rays, energy and timing resolution. We find that, while the pulse-shape discrimination is achievable with stilbene, further improvement of the readout is required to achieve it with the PSD-capable available plastic scintillators. [ABSTRACT FROM AUTHOR]

This paper reports a feasibility study of 235 U and 239 Pu characterization in 225 L bituminized waste drums or 200 L concrete waste drums, by detecting delayed fission gamma rays between the pulses of a deuterium-tritium neutron generator. The delayed gamma yields were first measured with bare samples of 235 U and 239 Pu in REGAIN, a facility dedicated to the assay of 118 L waste drums by Prompt Gamma Neutron Activation Analysis (PGNAA) at CEA Cadarache, France. Detectability in the waste drums is then assessed using the MCNPX model of MEDINA (Multi Element Detection based on Instrumental Neutron Activation), another PGNAA cell dedicated to 200 L drums at FZJ, Germany. For the bituminized waste drum, performances are severely hampered by the high gamma background due to 137 Cs, which requires the use of collimator and shield to avoid electronics saturation, these elements being very penalizing for the detection of the weak delayed gamma signal. However, for lower activity concrete drums, detection limits range from 10 to 290 g of 235 U or 239 Pu, depending on the delayed gamma rays of interest. These detection limits have been determined by using MCNPX to calculate the delayed gamma useful signal, and by measuring the experimental gamma background in MEDINA with a 200 L concrete drum mock-up. The performances could be significantly improved by using a higher interrogating neutron emission and an optimized experimental setup, which would allow characterizing nuclear materials in a wide range of low and medium activity waste packages. [ABSTRACT FROM AUTHOR]

Identification of neutron sources is of interest in several fields, such as homeland security or basic nuclear physics. Organic scintillators offer the possibility to unfold fast-neutron energy spectra. Since organic scintillators are sensitive to neutrons and gamma-rays, a method to discriminate these two types of particles is required. This paper is focused on investigation of the discrimination properties of the filter method by using a new procedure. We compare the results of an experimental filter method to a digital pulse-shape discrimination (PSD) method based on charge integration. In addition, these methods are compared to the simulation results obtained with the MCNPX-PoliMi code. The experimental and numerical investigations were performed with a 252 Cf spontaneous-fission neutron source. The integration of the number of counts gave the relative differences between the experimental filter and digital PSD neutron pulse-height distributions (PHDs) and simulated PHDs less than approximately 5% in the range between 60 keVee and 1.715 MeVee of light output. Above 1.715 MeVee, the PSD method has advantages over the filter method, due to the filter method having significantly worse counting statistics. The results show that the filter method has potential for robust neutron measurements when the PSD method cannot be applied, such as for ‘old’ organic plastic scintillators without PSD capability. [ABSTRACT FROM AUTHOR]

In this paper, we report on the application of Principal Component analysis (PCA) for pulse-shape discrimination (PSD) of scintillation radiation detectors. The details of the method are described and the performance of the method is experimentally examined by discriminating between neutrons and gamma-rays with a liquid scintillation detector in a mixed radiation field. The performance of the method is also compared against that of the conventional charge-comparison method, demonstrating the superior performance of the method particularly at low light output range. PCA analysis has the important advantage of automatic extraction of the pulse-shape characteristics which makes the PSD method directly applicable to various scintillation detectors without the need for the adjustment of a PSD parameter. [ABSTRACT FROM AUTHOR]

ENEA has been deeply involved in the design, development and testing of a laser based In Vessel Viewing System (IVVS) required for the inspection of ITER plasma-facing components. The IVVS probe shall be deployed into the vacuum vessel, providing high resolution images and metrology measurements to detect damages and possible erosion. ENEA already designed and manufactured an IVVS probe prototype based on a rad-hard concept and driven by commercial micro-step motors, which demonstrated satisfying viewing and metrology performances at room conditions. The probe sends a laser beam through a reflective rotating prism. By rotating the axes of the prism, the probe can scan all the environment points except those present in a shadow cone and the backscattered light signal is then processed to measure the intensity level (viewing) and the distance from the probe (metrology). During the last years, in order to meet all the ITER environmental conditions, such as high vacuum, gamma radiation lifetime dose up to 5 MGy, cumulative neutron fluence of about 2.3 × 10 17 n/cm 2 , temperature of 120 °C and magnetic field of 8 T, the probe mechanical design was significantly revised introducing a new actuating system based on piezo-ceramic actuators and improved with a new step focus system. The optical and mechanical schemes have been then modified and refined to meet also the geometrical constraints. The paper describes the mechanical concept design solutions adopted in order to fulfill IVVS probe functional performance requirements considering ITER working environment and geometrical constraints. [ABSTRACT FROM AUTHOR]

A piezoceramic motor has been identified as the potential apparatus for carrying out the rotation of the scanning head of a laser radar system used for viewing the first wall of the ITER vessel. This diagnostic is simply referred to as IVVS (In Vessel Viewing System). The choice fell on a piezoceramic motor due to the presence of strong magnetic fields (up 8 T) and of the high vacuum and temperature conditions. To be compliant with all the ITER environmental conditions it was necessary to qualify the piezo-motor under gamma and neutron irradiation. In this paper are described the procedures and tests that have been performed to verify the compatibility of the operation of the motor adopted in the presence of a fast neutron fluence which was gradually increased over time in order to reach a total value of 2.77 × 10 17 n/cm 2 . Such neutron fluence was obtained by irradiating the motor in a position close to the core of the fast nuclear reactor TAPIRO, in operation at the ENEA Casaccia Research Centre, Italy. The neutron spectrum in this position has been identified as representative of that found in the rest position of the IVVS head during ITER operation. The cumulative neutron fluence reached corresponds to that it is expected to be reached during the entire life of ITER for the IVVS in the rest position without any shield. This work describes the experimental results of this test; the methodology adopted to determine the total neutron fluence achieved and the methodology adopted for the verification of the correct operation of the motor. [ABSTRACT FROM AUTHOR]

Neutron howitzers are routinely used in universities to activate samples for instructional laboratory experiments on radioactivity. They are also a convenient source of neutrons and gammas for research purposes, but they must be used with caution. This paper describes the modeling, design, construction, and testing of a portable, economical shield for a 1.0 Curie neutron howitzer. The Monte Carlo N Particle Transport Code (MCNP5) has been used to model the 239 PuBe source and the howitzer and to design the external neutron and gamma shield. [ABSTRACT FROM AUTHOR]

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation. Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing. In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304–309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137 Cs source is also compared with the experimental data. [ABSTRACT FROM AUTHOR]

Neutron attenuation, scattering or radiative capture are used in various non-destructive methods to gain morphological, structural, elemental or isotopic information about the sample under study. The combined use of position-sensitive prompt gamma-ray detection (i.e. prompt gamma-ray activation imaging, PGAI) and neutron radiography/tomography (NR/NT) makes it possible to determine the 3D distribution of major elements and to visualize internal structures of heterogeneous objects in a non-destructive way. Based on earlier experience, the first ever permanent facility for this purpose, NIPS–NORMA, was constructed at the Budapest Neutron Centre, Hungary in 2012. The installation consists of a well-shielded, Compton-suppressed HPGe detector; a CCD-camera based imaging equipment and a motorized positioning system with sample support. Conventional PGAA measurements and NR/NT imaging using guided cold neutrons are the basic methods that form the basis of the more sophisticated experimental method called NR/NT-driven PGAI. The current status of the experimental station and its characteristics are described in the present paper. [ABSTRACT FROM AUTHOR]

The paper summarizes the current status of neutronics at ITER and a first set of proposals for experimental programmes to be conducted in the early operational life-time of ITER are described for the more crucial areas. These include a TF coils heating benchmark, a streaming benchmark and streaming measurements by activation on ITER itself. Also on ITER the measurement of activated water from triton burn-up should be planned and performed. This will require the measurement of triton burn-up in DD phase. Measurements of neutron flux in the tokamak building during DD operations should also be carried out. The use of JET for verification of shut down dose rate estimates is desirable. Other facilities to examine the production and behaviour of activated corrosion products and the shielding properties of concretes to high energy (6 MeV) gamma-rays are recommended. [ABSTRACT FROM AUTHOR]

Abstract: Lightning and thunderstorm systems in general have been recently recognized as powerful particle accelerators, capable of producing electrons, positrons, gamma-rays and neutrons with energies as high as several tens of MeV. In fact, these natural systems turn out to be the highest energy and most efficient natural particle accelerators on Earth. Terrestrial Gamma-ray Flashes (TGFs) are millisecond long, very intense bursts of gamma-rays and are one of the most intriguing manifestation of these natural accelerators. Only three currently operative missions are capable of detecting TGFs from space: the RHESSI, Fermi and AGILE satellites. In this paper we review the characteristics of TGFs, including energy spectrum, timing structure, beam geometry and correlation with lightning, and the basic principles of the associated production models. Then we focus on the recent AGILE discoveries concerning the high energy extension of the TGF spectrum up to 100MeV, which is difficult to reconcile with current theoretical models. [Copyright &y& Elsevier]

Abstract: Transport codes like MCNPX (Monte Carlo N-Particle eXtended) which are based primarily on experimentally derived data libraries at energies below approximately 100 MeV are reliable and predictive in transporting and producing neutrons and gammas. However, they are deficient in the microscopic sense because of the inability to produce correlated secondary particles. Neutrons will produce the average number of gammas at each collision regardless of the specific neutron reaction. Consequently, capture gammas may be produced by elastic collisions and inelastic gammas may be produced by capture. To remedy this deficiency the Cascading Gamma-ray and Multiplicity (CGM) V3.4 code has been integrated into MCNPX V2.7.0. The CGM code is based on Hauser–Feshbach and optical models. This paper details an overview of the CGM code integration into MCNPX followed by a test case highlighting the gamma spectrum and multiplicities from the CGM/MCNPX code. The test case provides gamma spectrum and multiplicities for the (n,γ) reaction on 56Fe using thermal neutrons. A second test case highlights gamma correlation for the (n,γ) reaction on 56Fe using thermal neutrons using three coincidence detectors. [Copyright &y& Elsevier]

Abstract: In this paper, using zero-crossing method, the neutron–gamma discrimination has been discussed to estimate the contribution of light transport on the timing characteristics of scintillation detectors. Both experimental and simulation verifications for the influence of lightguide length on the neutron–gamma discrimination quality of a 2×2 in. NE213 scintillator have been presented. [Copyright &y& Elsevier]

Abstract: We present an empirical model that describes the yield of gamma rays emitted by fission in the time interval from 20 to 958ns following a fission event. The analysis is based on experimental data from neutron-induced fission of 235U and 239Pu. The model is devised by first using regression analysis to identify likely patterns in the data and to choose plausible fitting functions. We provide statistical and physical arguments in support of time and energy independence. The intensity of the emitted gamma rays can be described as a bivariate distribution that is the product of independent variates for energy and time. We test several plausible distribution families for the energy and time variates and use maximum likelihood and minimum χ 2 to estimate distribution parameters. Because of the uncertainty in the experimental data, multiple combinations of variate pairs give rise to a surface that plausibly well fits the observations well. The best-fit variate turns out to be lognormal in energy and F in time. The findings illustrated in this paper can be used to simulate gamma ray de-excitation from fission in Monte Carlo codes. [Copyright &y& Elsevier]

Abstract: Phototransfered thermoluminescence (PTTL) found its use in radiation dosimetry for dose reassessment primarily in photon radiation fields. After the dose has first been evaluated by a “normal” thermoluminescence (TL) measurement, the same detector is subjected to the PTTL dose reassessment procedure providing a new estimation by an independent measurement procedure. PTTL also gives a new possibility for a dose reassessment and for separate dose determination in the mixed fields. It was shown in our previous papers that it is possible to determine separately gamma and neutron dose components with a single detector due to different sensitivities of TL and subsequent PTTL to neutrons in comparison to gamma rays. The main disadvantage, the low sensitivity of both, TL and PTTL, to neutrons restricts the use only for the case of radiation accident. In this work we have investigated the possibilities of improvement the sensitivities of PTTL by heating the samples during UV irradiation and by using the sensitization technique involving a “pre-dose” followed by annealing. Two types of LiF:Mg,Ti detectors, TLD-100 and TLD-700, were examined by irradiation in 137Cs gamma ray field and with neutrons of 14.5 MeV. After UV irradiation at different temperatures, PTTL increases reaching the values 2.1 and 2.5 times higher at 90 °C than that at 25 °C for TLD-100 and TLD-700 respectively. Increasing level is approximately the same for gamma rays and neutrons. After sensitization, TL signal is one half of that before sensitization. PTTL signal has increased 11 times for gamma rays and 5 times for neutrons at 25 °C with significant increase with increasing temperature during UV irradiation. [Copyright &y& Elsevier]

Abstract: The engineering design of fission chambers as on-line radiation detectors for IFMIF is being performed in the framework of the IFMIF-EVEDA works. In this paper the results of the experiments performed in the BR2 reactor during the phase-2 of the foreseen validation activities are addressed. Two detectors have been tested in a mixed neutron-gamma field with high neutron fluence and gamma absorbed dose rates, comparable with the expected values in the HFTM in IFMIF. Since the neutron spectra in all BR2 channels are dominated by the thermal neutron component, the detectors have been surrounded by a cylindrical gadolinium screen to cut the thermal neutron component, in order to get a more representative test for IFMIF conditions. The integrated gamma absorbed dose was about 4×1010 Gy and the fast neutron fluence (E >0.1MeV) 4×1020 n/cm2. The fission chambers were calibrated in three BR2 channels with different neutron-to-gamma ratio, and the long-term evolution of the signals was studied and compared with theoretical calculations. [Copyright &y& Elsevier]

Abstract: The 2D neutron emissivity profile is measured at JET by a set of NE213 liquid organic scintillators, located in two fan-shaped arrays of collimators with 10 horizontal and 9 vertical lines of sight. As the detectors are sensitive to neutron and gamma rays, pulse shape analysis is required for neutron/gamma discrimination. A digital architecture data-handling data acquisition and processing for the entire diagnostic has been developed by ENEA-Frascati and installed at JET during the 2009 campaign as a replacement of the existing analogue electronics. This paper describes the performances of the digital system and the first experimental results in the JET C27b campaign. [Copyright &y& Elsevier]

Abstract: Multiplicity counters for neutron assay have been extensively used in materials control and accountability for nonproliferation and nuclear safeguards. Typically, neutron coincidence counters are utilized in these fields. In this paper we present a measurement system that makes use not only of neutron (n) multiplicity counting but also of gamma ray multiplicity counting and the combined higher-order multiplets containing both neutrons and gamma rays. The benefit of this approach is in using both particle types available from the sample, leading to a reduction in measurement time when using more measurables. We present measurement results of n, , nn, , nnn, and multiplets emitted by a 252Cf source and a 239Pu–Be source. A dual radiation measuring system proposed here could use extra measurables either to improve the statistics when compared to a neutron-only system, or alternatively be used for extended analysis and interpretation of sample parameters. The study presented here provides an answer to the questions regarding the ability to measure both neutron and gamma ray coincidences at once. The results show that the measurements system proposed in this study has a potential to be valuable in the area of nuclear nonproliferation and homeland security. [ABSTRACT FROM AUTHOR]

Abstract: CUORE is a 1 ton scale cryogenic experiment aiming at the measurement of the Majorana mass of the electron neutrino. The detector is an array of 988 TeO2 bolometers used for a calorimetric detection of the two electrons emitted in the of 130Te. The sensitivity of the experiment to the lowest Majorana mass is determined by the rate of background events that can mimic a . In this paper, we investigate the contribution of external sources i.e. environmental gammas, neutrons and cosmic ray muons to the CUORE background and show that the shielding setup designed for CUORE guarantees a reduction of this external background down to a level <10−2 counts/kgkeVyr at the Q-value, as required by the physical goal of the experiment. [Copyright &y& Elsevier]

Abstract: BC-501 is a liquid scintillation detector sensitive to both neutrons and gamma rays. As these produce slightly different signals in the detector, they can be discriminated based on their pulse shape (Pulse Shape Discrimination, PSD). This paper reports on results obtained with several PSD techniques and compares them with a method based on artificial neural networks (NN) developed for this application. Results indicated a large performance advantage of NN especially in the region of small deposited energy which typically contains the majority of the events. NN were also applied for discrimination of pile-up events with good results. This framework can be implemented on some of the most recent programmable data acquisition cards and it is suitable for real-time application. [Copyright &y& Elsevier]

The Doppler broadening of gamma rays occurs in the fast neutron analysis with photons emitted in inelastic scattering reactions on certain nuclei, such as 12C and 14N. In this paper, we developed an algorithm based on the Monte Carlo method to calculate the Doppler broadened gamma peak shape of the 4.438 MeV gamma ray emitted by 12C in graphite when exposed to fast neutrons of different energies. The SAND-II unfolding iterative method is applied to obtain the spectrum of the incident neutron beam by using the Doppler broadened gamma spectra. In the calculation example, the unfolded neutron spectrum shows reasonable agreements with the original incident neutron spectrum. • Developed a Monte Carlo method to simulate the Doppler broadening process. • Calculated the Doppler broadening gamma peaks for different neutron energies. • Unfolded the neutron spectrum basing on the Doppler broadened gamma spectra. [ABSTRACT FROM AUTHOR]

Abstract: In this paper, we describe the implementation of a new source option in the Monte Carlo code MCNP-PoliMi. The code can be used to simulate time correlation measurements with neutrons and photons in plastic scintillation detectors. The new source describes the emission of neutrons and gamma-rays from an Am–Li (alpha,n) source. The simulations performed with the modified MCNP-PoliMi code are compared to corresponding measurements, and the reasons for disagreement are identified and discussed. [Copyright &y& Elsevier]

Abstract: Integrated tests (IT) applied to the Units I and II of the CNAAA Nuclear Power Plants (Angra dos Reis, Rio de Janeiro, Brazil) vara part of a comprehensive test program required to have the commercial operational licence issued. The IT cover operation at zero power and the subsequent power levels (the power escalation) until the nominal power operation. This paper reports and compares the results of systematic neutron and gamma equivalent dose-rate measurements performed inside and outside the reactor containment of each Unit, during the start-up phase of the operation, for different power levels of the reactors. It also compares the collective-doses measured along the operation of the two units of the CNAAA wich similar data of the Grafenrheinfeld nuclear power plant (Germany) — basic-of-project for the CNAAA — Unit II — and analyses the data under the fight of the state-of-art. [Copyright &y& Elsevier]

In the present paper, we address the issue of simulating nuclear safeguards measurements based on the time correlation between detections of fast neutrons and gamma rays by plastic scintillators. The Monte Carlo code MCNP-PoliMi is a modified version of MCNP that attempts to retain physical correlations in the fission process and in the gamma generation. An output file is generated that records the history of the single interactions inside the detector. Then, a post-processing code is used to analyze the data output and determine the detector response. We present some calibration experiments without the fissile sample aiming at verifying the effectiveness of both the modified code and post-processor in the description of the detector response to neutrons and gamma rays from the source. The post-processing code lends itself to analyzing the effect of varying relevant parameters, such as light output and pulse generation time, to generate the simulated detector response. By so doing, we determined the importance of the involved phenomena. We show that the correlation curves thereby obtained are in good agreement with the experiments. [Copyright &y& Elsevier]

In this paper, we demonstrate the application of a fast waveform digitizer (1 GHz) for neutron spectroscopy. We investigated pulse-shape discrimination between neutrons and γ-rays, timing properties and count rate capabilities of this new experimental method and compared the results with traditional techniques. The present method compares particularly well in the case of particle-type identification in the presence of pulse saturation and in case of high count rate. Excellent timing resolution and good pulse amplitude resolution is demonstrated. [Copyright &y& Elsevier]

This paper reports the first attempt to use L-band spectroscopy for estimating the sensitivity of whole teeth to fast neutrons and gamma-rays. Three teeth were successively irradiated first with fast neutrons with a wide energy spectrum (mean energy around 30 MeV) up to ∼160 Gy and then with gamma-rays up to ∼14 Gy. After each irradiation, L-band (∼1 GHz) EPR spectra of each whole tooth surrounded by the surface–coil resonator were recorded, yielding a single composite line principally due to CO2− and native radicals. The sensitivities are estimated by the slopes of the linear dose response curves of the dosimetric CO2− radicals. The ratios of the gamma/neutron sensitivities were found to be in the range 8–9 (±2) for the three teeth. [Copyright &y& Elsevier]

A neutron detector array consisting of a cylindrical plastic scintillator and 24 Helium-3 proportional counters is developed for the detection of neutrons with an extremely low rate. This detector array is designed for the JUNA (Jinping Underground Nuclear Astrophysics laboratory) project to measure the 13C (α , n) 16O reaction cross-section at stellar energies in CJPL (China Jinping Underground Laboratory). In this paper, we introduce the design of this detector array. The responses of the plastic scintillator to gamma rays and neutrons are tested at the ground level by shielding the detector array with a lead room. The neutron detection efficiency is obtained with a 252Cf source and a PuC source. The measured result is reproduced by Monte Carlo simulation. The coincidence measurement between plastic scintillator and Helium-3 counters is also performed and the coincidence times are obtained. These tests lay the foundation for the future measurement of the 13C (α , n) 16O reaction cross-section in CJPL. [ABSTRACT FROM AUTHOR]

In this paper, attempts are made to use waste plastics to make a novel Waste Polymer Incorporated Concrete (WPIC) mixes for gamma and neutron radiation shielding purposes. These mixes are achieved by simultaneous incorporation of waste polymeric material and high density fine and coarse aggregate components from iron industry in place of conventional ingredients. The results have indicated that the waste plastics along with by-products of iron industry can be conveniently used to make concrete mixes which have acceptable compressive strength characteristics and significantly enhanced shielding capabilities with respect to both gamma and neutron radiations. [Display omitted] • HDPE based waste polymeric material is used to make novel shielding concrete for gamma and neutron radiation shielding. • Mixes are achieved by simultaneous incorporation of waste polymeric material and high density fine and coarse aggregates. • Significant reductions in flux and dose transmissions and significant improvements in attenuation coefficients are observed. • Waste polymers with by-products of iron industry are used to make novel gamma and neutron radiation shielding concrete. [ABSTRACT FROM AUTHOR]

• A simple white spots removal method is proposed for degraded neutron images. • A'detection-location-removal' iterative strategy is introduced for the white spots detection and removal. • The qualitative and quantitative analysis are used for evaluating the performance of the mentioned methods. • The proposed method has a better performance in spots removal and details preservation than the comparison methods. Neutron radiography (NR) technology has been widely used in non-destructive investigations. However, the neutron images are always polluted by random white spots due to the interactions between the gamma rays, scattered neutron and other radiation and chip of the charge-coupled device (CCD) camera. In this paper, we propose a novel denoising method based on a 'detection-location-removal' iterative strategy, which adopts an improved robust principal component analysis (RPCA) in combination with the suitable threshold selection strategy and median filters for white spots noise removal of neutron images. Experimental results based on both simulation testing and real neutron images show that the proposed method is effective in improving neutron image quality in terms of numerical criteria and visual results. The comparison with the other denoising methods shows that the proposed method possesses higher robustness and effectiveness in practical denoising applications. [ABSTRACT FROM AUTHOR]