Analysis of microstress in neutron irradiated polyester fibre by X-ray diffraction technique

Microstresses developed in the crystallites of polymeric material due to irradiation of high-energy particle causes peak broadening and shifting of X-ray diffraction lines to lower angle. Neutron irradiation significantly changes the material properties by displacement of lattice atoms and the generation of helium and hydrogen by nuclear transmutation. Another important aspect of neutron irradiation is that the fast neutron can produce dense ionization at deep levels in the materials. The polyethylene terephthalate (PET) fibre of raw denier value, 78.2, were irradiated by fast neutron of energy, 4.44 MeV, at different fluences ranging from 1×10⁹ n/cm² to 1 × 10¹² n/cm². In the present work, the radiation heating microstresses developed in PET micro-crystallites was investigated applying X’Pert-MPD Philips Analytical X-ray diffractometer and the effects of microstresses in tensile strength of fibre measured by Instron have also been reported. The shift of 0.45 cm⁻¹ in the Raman peak position of 1614.65 cm{−1} to a higher value confirmed the development of microstresses due to neutron irradiation using micro-Raman technique. The defects due to irradiation were observed by SEM micrographs of single fibre for virgin and all irradiated samples.     

Degradation of cellulose nitrate with fast neutrons and gamma rays and their application in radiation dosimetry

Fast and moderated neutrons emitted from ²⁵²Cf as well as low doses of gamma rays from ⁶⁰Co, produce damaging effects in cellulose nitrate which can be determined viscometrically by calculating the average molecular weight at different doses.Samples were exposed to different doses of gamma rays (1 × 10^?4 to 1 Gy) and fission neutron fluences (10⁵–10¹¹ n/cm²) in free space and on a paraffin phantom. The effect of phantom thickness and phantom-to-detector distance on the detector readout have been investigated.The results revealed that the predominant bulk effects of radiation on CN is accelerated degradation by random chain scission.Empirical formulae have been given to calculate the absorbed doses of gamma rays and fast neutrons from the measured average molecular weight of the irradiated samples.     

A molecular fraction method for measuring personnel radiation doses

This work represents a development in fast and albedo neutron and gamma ray dosimetry, using cellulose nitrate, as a tissue equivalent material, in which radiation damage was registered.The changes in molecular fractions of the polymer were measured after irradiation with neutron fluences from a ²⁵²Cf source in the range 10⁵−10¹⁰ n/cm² and gamma doses in the range 10⁻⁴–10⁻¹ Gy through the use of gel filtration chromatography.Effects of irradiation on phantom, phantom to dosimeter distance, phantom thickness and storage at extreme environmental conditions were studied on the detector response and readout.The results showed that main chain scission followed by formation of new molecular configurations is the predominant effect of radiation on the polymer. The method enables measurements of neutron fluences and gamma doses in mixed radiation fields. Empirical formulae for calculating the absorbed dose from the measured changes in molecular fraction intensities are given.     

First Tests of 6Li Doped Glass Scintillators for Ultracold Neutron Detection

We report the results of test measurements aimed at determining the performances of ⁶Li doped glass scintillators for the detection of ultra-cold neutrons. Four types of scintillators, GS1, GS3, GS10 and GS20, which differ by their ⁶Li concentrations, have been tested. The signal to background separation is fully acceptable. The relative detection efficiencies have been determined as a function of the neutron velocity. We find that GS10 has a higher efficiency than the others for the detection of neutrons with velocities below 7 m/s. Two pieces of scintillators have been irradiated with a high flux of cold neutrons to test the radiation hardness of the glasses. No reduction in the pulse height has been observed up to an absorbed neutron dose of 1 × 10¹³ cm⁻³.     

Effect of neutron radiation on the photoelectric parameters of <Emphasis Type="Italic">p-n</Emphasis>-InSe structures

The effect of irradiation with fast reactor neutrons at an effective energy of 1 MeV and a fluence within Φ = 1 × 10¹⁴?5 × 10¹⁵ n/cm² on the photoelectric parameters of p-n-InSe homojunctions obtained in direct optical contact between p- and n-type semiconductors has been studied. The exposure to fast neutrons leads to an increase in the rectification coefficient and the diode ideality factor of the current-voltage characteristics with increasing neutron dose. No significant changes have been observed in the photosensitivity spectra of p-n-InSe structures irradiated to various doses, which allows these structures to be recommended for the creation of radiation-resistant photodetectors.     

Influence of neutron bombardment of diffused carbide and boride coatings on refractory metals

The effect of neutron bombardment on carbide and boride coatings on refractory metals was studied. It was found that a radiation dose of 10²⁰ neutrons/cm² produces cracking of the carbide coatings and exfoliation of the boride coatings. The radiation resistance of coatings of the kind studied depends to a large extent on the difference in the radiation-induced expansion of the coating and the substrate material.     

Development of fast neutron profiling method

We discuss the effects of neutron scattering and γ-ray background in fast neutron imaging and the method to reduce them. As a profiling device, a combination of an imaging plate (IP) and a polypropylene film (CH₂) has been employed in this study. Good profiles were obtained by employing appropriate neutron energy, a CH₂ thickness and geometry for accelerator-based fast neutrons (5–14 MeV). The neutron flux was ∼3.5×10⁴ cm⁻² s⁻¹ in the CH₂–IP position. Furthermore, we are designing the device using a position-sensitive photomultiplier in order to improve the signal-to-noise ratio by obtaining the information of pulse height for particle selection together with position.     

Neutron dosimetry using activation of thermoluminescent CaSO4

Sulfur activation in calcium sulfate doped with dysprosium (CaSO₄:Dy) thermoluminescent powder, which is bound in pure sulfur, has been used to measure the fast neutron dose at the tangential beam port of a Triga Mark III reactor. After a post-irradiation time of 3 d, the dosimeters were annealed at 600°C for 30 min in order to erase all the thermoluminescence acquired during the irradiation. The dosimeters were then stored to allow self-irradiation by betas from ³²P produced by sulfur activation. The thermoluminescent signal accumulated during a post-irradiation time of 20 d due to a neutron fluence of 2.2 × 10¹¹ n/cm² was equivalent to an absorbed dose of 10 mGy of ⁶⁰Co gamma rays. The thermoluminescence as a function of fast neutron dose fitted to a straight line on a log-log scale from 1 Gy to 10⁴Gy.     

Electrical properties of ion irradiated polypropylene films

The effect of high-energy (50 MeV) Li³⁺ ion beam irradiation on polypropylene (PP) film has been studied in the fluence range 2.4 × 10¹²−l.5 × 10¹⁴ ions/cm². The a.c. electrical properties of PP films were measured in the frequency range from 0.05– 100 kHz, and at temperature range between 30 and 140°C. This study indicates two peaks at 60°C and 120°C with comparatively high magnitudes. There is an exponential increase in conductivity with log of frequency and the effect is significant at higher fluences. The loss factor (tan δ) vs frequency plot suggests that PP film based capacitors may be useful below 10 kHz. The capacitance is constant over a wide temperature range up to 130°C. FTIR spectra of the PP films before and after irradiation indicate that intensity of C-H stretching vibration at 2900 cm⁻¹ is modified. The presence of many new peaks with the increase of fluence suggests the formation of alkanes and alkynes which might be responsible for the observed changes in the dielectric and electrical properties of PP films.     

NECTAR—A fission neutron radiography and tomography facility

NECTAR (Neutron Computerized Tomography and Radiography) is a versatile facility for radiographic and tomographic investigations as well as for neutron activation experiments using fission neutrons.The radiation sources for this facility are two plates of highly enriched uranium situated in the moderator vessel in FRM II. Thermal neutrons originating from the main fuel element of the reactor generate in these plates fast neutrons. These can escape through a horizontal beam tube without moderation. The beam can be filtered and manipulated in order to reduce the accompanying gamma radiation and to match the specific experimental tasks.A summary of the main parameters required for experimental set-up and (quantitative) data evaluation is presented. The (measured) spectra of the neutron and gamma radiations are shown along with the effect of different filters on their behavior. The neutron and gamma fluxes, dose rates, L/D-ratios, etc. and the main parameters of the actually used detection systems for neutron imaging are given, too.     

He and BF3 neutron detector pressure effect and model comparison

Radiation detection systems for homeland security applications must possess the capability of detecting both gamma rays and neutrons. The radiation portal monitor systems that are currently deployed use a plastic scintillator for detecting gamma rays and ³He gas-filled proportional counters for detecting neutrons. Proportional counters filled with ³He are the preferred neutron detectors for use in radiation portal monitor systems because ³He has a large neutron cross-section, is relatively insensitive to gamma-rays, is neither toxic nor corrosive, can withstand extreme environments, and can be operated at a lower voltage than some of the alternative proportional counters. The amount of ³He required for homeland security and science applications has depleted the world supply and there is no longer enough available to fill the demand. Thus, alternative neutron detectors are being explored.Two possible temporary solutions that could be utilized while a more permanent solution is being identified are reducing the ³He pressure in the proportional counters and using boron trifluoride gas-filled proportional counters. Reducing the amount of ³He required in each of the proportional counters would decrease the rate at which ³He is being used; not enough to solve the shortage, but perhaps enough to increase the amount of time available to find a working replacement. Boron trifluoride is not appropriate for all situations as these detectors are less sensitive than ³He, boron trifluoride gas is corrosive, and a much higher voltage is required than what is used with ³He detectors. Measurements of the neutron detection efficiency of ³He and boron trifluoride as a function of tube pressure were made. The experimental results were also used to validate models of the radiation portal monitor systems.     

Dose measuring system for boron neutron capture therapy

Dose measuring systems for boron neutron capture therapy (BNCT) of brain tumors are presented. The systems are a real-time monitoring system, an integral measuring system and a ¹⁰B concentration measuring system. The real-time monitoring with a small PN junction silicon detector made it possible to simultaneously measure the thermal neutron flux and the gamma dose rate in a patient during neutron therapy. Another monitoring of dose equivalents of thermal neutrons and gamma rays was performed with a BGO scintillation detector connected to an optical fiber. The accurate neutron fluence and gamma dose were determined with the integral measurements of the foil activation method and thermoluminescent dosimeters (TLDs) after irradiation. Kerma doses of thermal neutrons and gamma-rays were also measured with the TLD at the same time. Preliminary measurements of ¹⁰B concentration in tissue and blood of a patient were carried out by prompt gamma-ray spectroscopy.     

Neutronic design and simulated performance of Peking University Neutron Imaging Facility (PKUNIFTY)

The Peking University Neutron Imaging Facility (PKUNIFTY) is a Radio Frequency Quadruple (RFQ) accelerator based system. The fast neutrons are produced by 2 MeV deuterons bombarding beryllium target. The moderator, reflector, shielding and collimator have been optimized with Monte-Carlo simulation to improve the neutron beam quality. The neutrons are thermalized in water cylinder of Φ26×26 cm² with a polyethylene disk in front of Be target. The size of deuteron beam spot is optimized considering both the thermal neutron distribution and the demand of target cooling. The shielding is a combination of 8 cm thick lead and 42 cm thick boron doped polyethylene. The thermal neutrons are extracted through a rectangular inner collimator and a divergent outer collimator. The thermal neutron beam axis is perpendicular to the D⁺ beam line in order to reduce the fast neutron and the γ ray components in the imaging beam. When the neutron yield is 3×10¹² n/s and the L/D is 50, the thermal neutron flux is 5×10⁵ n/cm²/s at the imaging plane, the Cd ratio is 1.63 and the n/γ ratio is 1.6×10¹⁰ n/cm²/Sv.     

GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities.In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]).The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10¹¹ ncm²/s.This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted.The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line (http://www.info.cern.ch/asd/geant4/geant4.html[4]).To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al₂O₃) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al₂O₃) filters, respectively.To get a good cadmium ratio, GEANT 4 simulations were used to define the design of the moderator in the inlet of the radiation channel. A graphite block of 22 cm thickness seems to be the optimal neutron moderator.The results showed that the combination of 5 cm of bismuth with 5 cm of sapphire permits the filtration of gamma-rays, epithermal neutrons as well as fast neutrons in a considerable way without affecting the neutron thermal flux.     

Development of advanced-type multi-functional electronic personal dosemeter

An advanced-type small, light, multi-functional electronic personal dosemeter has been developed using silicon semiconductor radiation detectors for dose management of workers at nuclear power plants and accelerator facilities. This dosemeter is 62 x 82 x 27 mm(3) in size and approximately 130 g in weight, which is capable of measuring personal gamma ray and neutron dose equivalents, Hp(10), simultaneously. The neutron dose equivalent can be obtained using two types of silicon semiconductors: a slow-neutron sensor (<1 MeV) and a fast-neutron sensor (>1 MeV). The slow neutron sensor is a 10 x 10 mm(2) p-type silicon on which a natural boron layer is deposited around an aluminium electrode. The fast neutron sensor is also a 10 x 10 mm(2) p-type silicon crystal on which an amorphous silicon hydride is deposited. The neutron energy response corresponding to the fluence-to-dose-equivalent conversion coefficient given by ICRP Publication 74 has been evaluated using a monoenergetic neutron source from 250 keV to 15 MeV at the Fast Neutron Laboratory of Tohoku University. As the result, the Hp(10) response to neutrons in the energy range of 250 keV and 4.4 MeV within +/-50% difference has been obtained.     

Neutron field produced by 25 MeV deuteron on thick beryllium for radiobiological study; energy spectrum

Biological data is necessary for estimation of protection from neutrons, but there is a lack of data on biological effects of neutrons for radiation protection. Radiological study on fast neutrons has been done at the National Institute of Radiological Sciences. An intense neutron source has been produced by 25 MeV deuterons on a thick beryllium target. The neutron energy spectrum, which is essential for neutron energy deposition calculation, was measured from thermal to maximum energy range by using an organic liquid scintillator and multi-sphere moderated 3He proportional counters. The spectrum of the gamma rays accompanying the neutron beam was measured simultaneously with the neutron spectrum using the organic liquid scintillator. The transmission by the shield of the spurious neutrons originating from the target was measured to be less than 1% by using the organic liquid scintillator placed behind the collimator. The measured neutron energy spectrum is useful in dose calculations for radiobiology studies.     

Proton induced modification in makrofol-DE

Irradiation effects of a 3 MeV proton beam on polycarbonate (makrofol-DE (MFD)) have been studied with respect to its electrical, thermal and structural behaviour by using an LCR meter, DSC/TGA and FTIR spectroscopy. The dielectric loss/constant was observed to change with the fluence. Thermal analysis revealed that chain scission is the dominant phenomena in irradiated samples based on the reduction of its thermal stability by about 19% at a fluence of 10¹⁵ ions/cm², which is also corroborated by FTIR spectra. No significant change in intensity of the absorbance bands of the irradiated sample was observed up to a fluence of 10¹⁴ ions/cm² while on increasing fluence (10¹⁵ ions/cm²) the polymer structure was modified. It appears from DSC thermograms thatT _g is observed to change with fluence.     

Multiferroic properties of BiFeO3/Bi4Ti3O12 double-layered thin films fabricated by chemical solution deposition

Multiferroic BiFeO₃/Bi₄Ti₃O₁₂ (BFO/BTO) double-layered film was fabricated on a Pt(111)/Ti/SiO₂/Si(100) substrate by a chemical solution deposition method. The effect of an interfacial BTO layer on electrical and magnetic properties of BFO was investigated by comparing those of pure BFO and BTO films prepared by the same condition. The X-ray diffraction result showed that no additional phase was formed in the double-layered film, except BFO and BTO phases. The remnant polarization (2P_r) of the double-layered film capacitor was 100 μC/cm² at 250 kV/cm, which is much larger than that of the pure BFO film capacitor. The magnetization-magnetic field hysteresis loop revealed weak ferromagnetic response with remnant magnetization (2M_r) of 0.4 kA/m. The values of dielectric constant and dielectric loss of the double-layered film capacitor were 240 and 0.03 at 100 kHz, respectively. Leakage current density measured from the double-layered film capacitor was 6.1 × 10^− 7 A/cm² at 50 kV/cm, which is lower than the pure BFO and BTO film capacitors.     

Radiation-induced changes on some physical parameters of gelatine for protection-level dosimetry utilization

The induced effects of gamma and thermal neutrons of respective absorbed dose ranges 10^–2–10³ Gy and 10–500 mSv, on the physical changes of the gelatine electrical resistivity, hardness, and luminescence intensity (RPL), have been investigated. The results indicated that the hardness indentation number and RPL luminescence emission intensity increase with increasing gamma and neutron absorbed doses, while the electrical resistivity decreases. The dose sensitivity relation with these induced physical changes is beneficial for dosimetric utilization. The changes are described by semi-empirical regressions formulae.     

Neutron detection gamma ray sensitivity criteria

The shortage of ³He has triggered the search for effective alternative neutron detection technologies for national security and safeguards applications. Any new detection technology must satisfy two basic criteria: (1) it must meet a neutron detection efficiency requirement, and (2) it must be insensitive to gamma-ray interference at a prescribed level, while still meeting the neutron detection requirement. It is the purpose of this paper to define measureable gamma ray sensitivity criteria for neutron detectors. Quantitative requirements are specified for: intrinsic gamma ray detection efficiency and gamma ray absolute rejection. The gamma absolute rejection ratio for neutrons (GARRn) is defined, and it is proposed that the requirement for neutron detection be 0.9<GARRn<1.1 at a 10 mR/h exposure rate. An example of the results from a ³He based neutron detector is provided showing that this technology can meet the stated requirements. Results from tests of some alternative technologies are also reported.