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To compare the response to high energy neutrons and protons of the space bubble detectors in use aboard the International Space Station (ISS), three series of experiments were conducted with high energy protons and neutrons. The first series of experiments was conducted with high-energy neutrons in the energy range expected for neutrons encountered during space flight (0.6–800 MeV) at the Los Alamos Neutron Science Center (LANSCE) using the spallation neutron source. The second series was conducted with high energy protons from 30 to 70 MeV using the cyclotron at the National Institute of Radiological Science NIRS in Japan, and the third series of experiments was performed with high energy protons from 60 to 230 MeV at the ProCure proton therapy facility, Oklahoma, USA. The bubble detectors were exposed to different fluences in different experiments and the number of bubbles was counted using a bubble detector reader. The proton response of the bubble detector (sensitivity), as a function of energy, was determined and compared to the neutron sensitivity. In addition, to adjust the neutron sensitivity of the bubble detector determined in an AmBe field, a calibration factor was obtained for space applications.

Lanthanum Halide scintillators such as lanthanum bromide (LaBr3:Ce) and lanthanum chloride (LaCl3:Ce) have been studied extensively in gamma radiation fields and have shown to be excellent gamma ray detectors. Measurements with these detectors in complex radiation fields that include neutrons, protons and heavy ions may produce some information about the radiation field. For example fast neutron (n,γ), (n,p), and (n,α) reactions produce scintillations in the crystal from the energy deposited by the resulting secondary particles. Also high energy radiation environments such as those encountered in spacecraft and high energy charged particle facilities contain protons and heavy ions with enough energy to penetrate typical scintillation detectors. The light produced in the crystal by these heavy ions is proportional to the energy deposited along the ion track. To investigate the response of LaBr3 detectors to heavy ions, a series of experiments have been carried out at the Heavy Ion Medical Accelerator in Chiba, Japan (HIMAC). Measurements with He and Si ions with various incident energies have been performed and the light output of the scintillator has been measured. The experimentally measured spectra have been compared to simulations using the particle and heavy ion transport system (PHITS) code and quenching effects have been observed in the measured spectra. The quenching of the Si ion experiments is greater than the quenching observed in the He ion experiments. The results from both simulation and experiment are presented and discussed.

It is well known that electrical-power generation plays the key role in advances in industry, agriculture, technology, and standard of living. Also, strong power industry with diverse energy sources is very important for a country's independence. In general, electrical energy can be mainly generated from: (1) nonrenewable energy sources (75.5% of the total electricity generation) such as coal (38.3%), natural gas (23.1%), oil (3.7%), and nuclear (10.4%); and (2) renewable energy sources (24.5%) such as hydro, biomass, wind, geothermal, solar, and marine power. Today, the main sources for electrical-energy generation are: (1) thermal power (61.4%)—primarily using coal and secondarily using natural gas; (2) “large” hydro-electric plants (16.6%); and (3) nuclear power (10.4%). The balance of the energy sources (11.6%) is from using oil, biomass, wind, geothermal, and solar, and has visible impact just in a few countries. This paper presents the current status of electricity generation in the world, various sources of industrial electricity generation and role of nuclear power with a comparison of nuclear-energy systems to other energy systems. A comparison of the latest data on electricity generation with those several years old shows that world usage of coal, gas, nuclear, and oil has decreased by 1–2%, but usage of renewables has increased by 1% for hydro and 2% for other renewable sources. Unfortunately, within last years, electricity generation with nuclear power has decreased from 14% before the Fukushima Nuclear Power Plant (NPP) severe accident in March 2011 to about 10%. Therefore, it is important to evaluate current status of nuclear-power industry and to make projections on near (5–10 yr) and far away (10–25 yr and beyond) future trends.

Currently, nuclear power plays a quite visible role in the world electricity generation (∼11%). However, before the Fukushima Nuclear Power Plant (NPP) severe accident in March of 2011, NPPs generated about 14% of the world’s electricity. Accounting that after, mainly, Chernobyl NPP severe accident a number of power reactors built and put into operation in the world decreased from 120 within 1985–1990 to about 22 per 5 years (within 1995–2015), we might face a significant shortage of operating power reactors within 2030–2040. Therefore, it is important to evaluate current status of nuclear-power industry and to make projections on near (5–10 years) and far away (10–25 years and beyond) future trends in nuclear-power industry. In the current paper statistics on all current nuclear-power reactors were analyzed and based on that future trends were estimated in terms of types of reactors to be left after 10 years, new types of reactors to be put into operation, projections of how many reactors and of which types will be build. To make any projections an average operating term of power reactors should be estimated. In the current paper a nuclear-power-reactor operating term of 45 years was considered. Also, rates of building and putting into operation power reactors worldwide were estimated, and several scenarios of future developments in nuclear-power industry in the world and in selected countries were considered.

There are several heavy materials used as neutron spallation targets. The purpose of this study is to optimize the geometry of the neutron production target in the spallation process. using Monte-Carlo method-Particle code MCNP-6. The simulation includes Uranium, Thorium, and Tantalum targets. We investigated the flux of neutron produced as well as the neutron axial distribution by variation the spallation target, target height and proton beam energy. The proton beams of a 2 cm radius were assumed uniformly distributed.

A recently developed scintillator has been investigated for possible use as a dual detector for neutron and gamma spectrometry. A 7Li-enriched version of the scintillator has been investigated. The 35 Cl(n,p) 35 S nuclear reaction provides a possibility for fast neutron detection. The sensor has been mounted on a photomultiplier tube controlled with a miniature electronics board and irradiated in different gamma and neutron radiation fields. A series of experiments has been carried out with different gamma energies as well as well with mono-energetic neutrons from a KN Van de Graaff accelerator, and the pulse height spectra have been measured. To clarify different features observed on the response functions of the detector, a Monte Carlo model of the scintillator has been built using MCNP6 and emitted charged particles have been tracked. The simulation data along with the experiments are analyzed, compared and reported.

This paper presents a method for a small, inexpensive mobile robot equipped with a single high resolution scintillation detector to quickly survey an area and convey information about local sources of gamma radiation to a remote human operator. This is achieved by surrounding the detector with a lead sheath that blocks all gamma-rays except those incident along the detector's axial direction. A 180° horizontal scan is performed by rotating the detector and a directional profile of gamma radiation is constructed. In addition, a 180° visual panorama of the local area is assembled using a camera mounted on the detector. A plot of the detector count rate versus angle is then overlaid on top of the visual panorama and visible peaks clearly indicate the direction of local gamma radiation sources. Measuring the energy spectrum of gamma-rays in each direction produces a 2D count frequency histogram where distinct peaks indicate the energy and direction of local gamma-ray sources allowing the identification of different radio-isotopes. The mobile robot can use the peaks as goal directions and autonomously move towards gamma-ray sources.

Research into advanced screening technologies has become high priority in all aspects of occupational nuclear safety and environmental radiation protection. Neutrons are a fundamental part of radiation encountered in various fields of nuclear science and technology and their detection is still employing detectors with a high thermal neutron response embedded in a thermalizing medium where helium based devices have been a dominant choice in many applications. Recently, there have been newly developed sensors based on multi-elements that include 6 Li and 35 Cl isotopes to detect neutrons and gamma radiation. Among these new sensors one can cite the elpasolite scintillator, known as CLYC. This sensor contains two neutron sensitive isotopes and may serve as a dual detector for gamma as well as for neutron radiation. In this paper, the response functions of this sensor have been investigated in different fields of neutron and gamma-radiation. The sensor responses have been simulated using Monte Carlo N-Particle MCNPX code and a series of experiments have been carried out to validate the simulated data. Both sets of data are presented and discussed.

Measurements using bubble detectors have been performed in order to characterise the neutron dose and energy spectrum in the Russian segment of the International Space Station (ISS). Experiments using bubble dosemeters and a bubble-detector spectrometer, a set of six detectors with different energy thresholds that is used to determine the neutron spectrum, were performed during the ISS-22 (2009) to ISS-33 (2012) missions. The spectrometric measurements are in good agreement with earlier data, exhibiting expected features of the neutron energy spectrum in space. Experiments using a hydrogenous radiation shield show that the neutron dose can be reduced by shielding, with a reduction similar to that determined in earlier measurements using bubble detectors. The bubble-detector data are compared with measurements performed on the ISS using other instruments and are correlated with potential influencing factors such as the ISS altitude and the solar activity. Surprisingly, these influences do not seem to have a strong effect on the neutron dose or energy spectrum inside the ISS.

Monte Carlo simulation of a detector response function presents a very challenging problem. The detector response functions have been calculated for different neutron and gamma detectors: 3 He gas filled proportional counter, NE213 organic scintillator, BrillanCe 350 or LaCl 3 (Tl), and an ionization chamber with mixed gas composition. MCNPX code was used for simulations. The simulations were done with different neutron and gamma energies. The effects of neutron scattering, wall effects, recoil continua and contribution from charged particles have been included. The detector response function for the NE213 organic scintillator was obtained with consideration of light output curves of different products of neutron reactions with materials of the scintillator. The simulated data has been compared with experiments.

Rotating Spectrometer (ROSPEC) is a neutron spectrometer designed to measure neutron energy distributions, and provide accurate neutron dosimetry. It is a completely self-contained unit and measures neutron energy via recoiling protons in gas proportional counters. Each of the four original gas counters is dedicated to a particular neutron energy range dictated by sensitivity to gamma rays at the low energy end of the spectrum and by proton collisions with the counter walls at the high energy end. Introduced originally in 1992, ROSPEC has a proven operational record with a program of continued upgrades. The operating range of the original ROSPEC spans 50 keV-4.5 MeV. The range of the ROSPEC has now been extended down to include epithermal and thermal neutrons by adding two 2 in. (3)He counters. Also, an optional simple scintillation spectrometer was designed to extend the upper limit of ROSPEC up to 18 MeV.

This paper presents a novel method for making an inexpensive, low resolution, wide angle gamma-ray camera using a single handheld inorganic scintillation detector. A lead aperture is used to create a directionally sensitive detector and craterisation points are measured and compiled into a pixel Ted spherical gamma-ray image. The gamma-ray image is separated into energy channels and overlaid on top of a grey scale visual panoramic image so that collared pixels indicate the position and energy of gamma-ray sources.

Bubble detectors have been used to characterise the neutron dose and energy spectrum in several modules of the International Space Station (ISS) as part of an ongoing radiation survey. A series of experiments was performed during the ISS-34, ISS-35, ISS-36 and ISS-37 missions between December 2012 and October 2013. The Radi-N2 experiment, a repeat of the 2009 Radi-N investigation, included measurements in four modules of the US orbital segment: Columbus, the Japanese experiment module, the US laboratory and Node 2. The Radi-N2 dose and spectral measurements are not significantly different from the Radi-N results collected in the same ISS locations, despite the large difference in solar activity between 2009 and 2013. Parallel experiments using a second set of detectors in the Russian segment of the ISS included the first characterisation of the neutron spectrum inside the tissue-equivalent Matroshka-R phantom. These data suggest that the dose inside the phantom is ∼70 % of the dose at its surface, while the spectrum inside the phantom contains a larger fraction of high-energy neutrons than the spectrum outside the phantom. The phantom results are supported by Monte Carlo simulations that provide good agreement with the empirical data.

The neutron total cross-section of natural Dy has been measured in the energy region from 0.025 eV to 100 eV by the neutron time-of-flight method at Pohang Neutron Facility, which consists of an electron linear accelerator, a water-cooled Ta target with a water moderator, and a time-of-flight path with an 11 m length. A 6Li–ZnS(Ag) scintillator with a diameter of 12.5 cm and a thickness of 1.5 cm has been used as a neutron detector and a metallic plate of Dy sample have been used for the neutron transmission measurement. The background level has been determined by using notch-filters of Co, Ta, and Cd sheets. In order to reduce the gamma rays from a Bremsstrahlung and that from a neutron capture, we have employed a neutron-gamma separation system based on their different pulse shape. The present measurements are in general agreement with the previous ones and the evaluated data in ENDF/B-VI. The resonance parameters of Dy isotopes (161,162,163Dy) have been extracted from the transmission data by using the SAMMY code and compared with the previous ones.

The paper presents the creation of a simulator for the Geiger Muller (GM) gas filling station. It simulates the entire gas filling process, from preparing for gas filling by vacuuming out the system to adding in different gases and sealing the detector. The simulator not only demonstrates the overall gas filling process, but also identifies certain steps that can be automated to speed up the process and reduce human error. An easy-to-use human-machine interface is created so that the operator is informed of the current state of the system and the operations that need to be performed manually. A lab scale bench top test is set up to demonstrate the implementation of the automated process. It shows that a Compact RIO system running on real-time environment can be used to communicate with the computer and control a motor. Finally, in order to integrate the simulator with the current gas filling station, the improvements that need to be made on the current station are identified.

A new scintillator, CLYC, has been investigated for possible use in neutron spectrometry. This sensor provides neutron detection for both thermal and fast neutrons from the reactions 6Li(n,α) and 35Cl(n,p), respectively. This work primarily focuses on the detection of fast neutrons since there is currently no sensor that can accurately and efficiently provide information about their incident neutron energy. Conventional methods of fast neutron detection have been based on utilizing materials that use the elastic scattering process of neutrons on 1H to create recoil protons or by thermalizing and capturing these neutrons at thermal energies. Both approaches have drawbacks and are complex in deriving the energy spectrum through the unfolding process. The CLYC scintillator uses a distinct proton peak, whereby the position on the spectrum is proportional to the energy of the incident neutron. The response function of this detector has been simulated using Monte Carlo N-Particle eXtended code (MCNPX) for gamma-rays and neutrons of different energies. The obtained data has been discussed and analyzed.

A neutron facility at the University of Ontario Institute of Technology (UOIT) is described. Radiation fields around a DT neutron generator have been simulated. Spectral distributions of neutron emissions as well as radiation dose rates have been calculated for a neutron generator of 1.1010 n/s intensity and 14.2 MeV energy. The simulations have been performed at different locations around the neutron generator inside and outside a heavily shielded room. Additional shielding composition around the generator has been optimized to meet the regulatory requirements. A modern Monte Carlo code (MCNPX 2.7.E) has been used for simulations. Special features of the MCNPX code as variant reduction techniques to improve the simulation quality have been used and described. A linear motion system to control the position of the generator is also described.Copyright © 2012 by ASME

A boron-loaded plastic scintillator has been investigated for possible use in neutron spectrometry. The sensor composition of hydrogen and carbon leads to multiple scattering collisions that are useful for fast neutron spectroscopy, while its boron component can serve as a thermal neutron detector. Both simulation and experimental work have been carried out to investigate the response functions of the detector to neutrons and gamma radiation. The response functions of the detector have been simulated using Monte Carlo N-Particle eXtended code. For experimental tests, the sensor has been mounted on a photomultiplier tube connected to a compact data acquisition system. The system has been tested in different gamma and neutron fields at the University of Ontario Institute of Technology Neutron Facility. The simulation and experimental results have been compared and analyzed.Copyright © 2012 by ASME

A gamma spectrometer based on LaBr3(Ce) crystal has been developed. The spectrometer uses a compact data acquisition system including a multi-channel analyzer and corresponding electronics components. The response function of the detector has been simulated using Monte Carlo Code (MCNPX 2.7E). It has been also tested, at the University of Ontario Institute of Technology radiation facilities, using a series of gamma radiation sources. The simulated data has been compared with experiments and a good agreement has been achieved. This paper presents the parameters of the developed spectrometer and the results of its testing along with a comparison with NaI(TI) crystal.Copyright © 2012 by ASME

A passive neutron-bubble dosemeter (BD), developed by Bubble Technology Industries, has been used for space applications. Both the bubble detector-personal neutron dosemeter and bubble detector spectrometer have been studied at ground-based facilities in order to characterise their response due to neutrons, heavy ion particles and protons. This technology was first used during the Canadian-Russian collaboration aboard the Russian satellite BION-9, and subsequently on other space missions, including later BION satellites, the space transportation system, Russian MIR space station and International Space Station. This paper provides an overview of the experiments that have been performed for both ground-based and space studies in an effort to characterise the response of these detectors to various particle types in low earth orbit and presents results from the various space investigations.

As part of the Matroshka-R experiments, a spherical phantom and space bubble detectors (SBDs) were used on board the International Space Station to characterise the neutron radiation field. Seven experimental sessions with SBDs were carried out during expeditions ISS-13, ISS-14 and ISS-15. The detectors were positioned at various places throughout the Space Station, in order to determine dose variations with location and on/in the phantom in order to establish the relationship between the neutron dose measured externally to the body and the dose received internally. Experimental data on/in the phantom and at different locations are presented.

Tests of existing electronic neutron dosimeters by military and civilian groups have revealed significant performance limitations. To meet the operational requirements of emergency response personnel to a radiological/nuclear incident as well as those in the nuclear industry, a new END has been developed. It is patterned after a unique commercial neutron spectral dosemeter known as the N-probe.* It uses a pair of small special scintillators on tiny photomultiplier tubes. Special electronics were designed to minimize power consumption to allow for weeks of operation on a single charge. The size, performance, and data analysis for the END have been designed to meet/exceed international standards for electronic neutron dosimeters. Results obtained with the END prototype are presented.

Bubble Technology Industries (BTI) has built a revolutionary portable neutron scintillation spectrometer, N-Probe, designed to be used by non-specialists for measurement of low-intensity neutron doses in the mixed field environments often found in nuclear utilities, fuel storage areas, fuel and waste processing operations and military applications. It is compatible with the current generation of BTI MICROSPEC analysers and shares the philosophy of spectral dosimetry with other BTI spectroscopic probes, where the dosimetric quantities are computed from the spectrum using appropriate fluence-dose conversion functions.