Investigation of fast neutron shielding characteristics depending on boron percentages of MgB<Subscript>2</Subscript>, NaBH<Subscript>4</Subscript> and KBH<Subscript>4</Subscript>

The macroscopic cross-section Σ and average neutron fluence in matter Φ are usable factors to comment neutron shielding property of samples. In this paper, we have used MgB₂, NaBH₄ and KBH₄ samples including different percentages of boron. Neutron macroscopic cross-section measurements of them have been done by using a source of mono-energetic neutrons (E _eff = 4.5 MeV ²⁴¹Am–Be). Average neutron fluence values and double differential fast neutron flux distributions of each samples calculated by using FLUKA Monte Carlo code. Also half value layers (HVLs) of samples are compared to paraffin which is one of the most neutron moderators. As a result, growing boron concentration can raise neutron shielding property of materials.     

Activation rates and chemical recovery of67Cu produced with low energy proton irradiation of enriched70Zn targets

Copper-67 is a radioisotope with significant potential for diagnostic and therapeutic applications in nuclear medicine. Despite its promise,⁶⁷Cu has failed to make an impact in clinical nuclear medicine, primarily because it is available sporadically, and in limited quantities. Common methods of production rely on high energy proton irradiation of natural zinc targets or on induced reactions using high energy neutrons at nuclear reactors. We have evaluated alternative production methods that could provide year-round adequate supply of this isotope. Using a low energy accelerator, we have studied the production of⁶⁷Cu by proton reactions on enriched⁷⁰Zn. Our results indicate that it is possible to produce useful quantities of⁶⁷Cu from the irradiation of enriched⁷⁰Zn with protons that have energies of less than 20 MeV. Production rates are higher than currently used methods at high energy accelerators or reactors. This isotope can be made available throughout the year as a result of this research.     

Measurement of <Superscript>99</Superscript>Mo production cross-section from the <Superscript>100</Superscript>Mo(n,2n) reaction with quasi monoenergetic neutron based on the <Superscript>9</Superscript>Be(p,n) reaction

The production cross-section of the medical isotope, ⁹⁹Mo from the enriched ¹⁰⁰Mo(n,2n) reaction with the average neutron energies of 21.9 and 26.5 MeV has been determined for the first time by using an off-line γ-ray spectrometric technique. The average neutron energies were generated by using the ⁹Be(p,n) reaction with the proton energies of 35 and 45 MeV from the MC50 cyclotron of the Korea Institute of Radiological and Medical Sciences (KIRAMS) at Seoul, South Korea. The ¹⁰⁰Mo(n,2n) reaction cross-section as a function of neutron energy was also calculated theoretically by using the computer code TALYS-1.8 and EMPIRE-3.2 Malta. The experimental results are in close agreement with the theoretical values from TALYS-1.8. However, the present data at the neutron energy of 21.9 MeV is slightly lower and at 26.5 MeV is higher than the values from EMPIRE-3.2 Malta.     

A study of a transportable thermal neutron radiography unit based on a compact RFI linac

A transportable thermal neutron radiography system, incorporating a compact proton accelerator as neutron source has been simulated using the MCNP4B code. The neutron source will be produced via the ⁷Li(p,n)⁷Be reactions by a 2.5?MeV, 10?mA proton beam into a thick lithium target. Variable values for the collimator ratio were calculated. Thermal neutron radiography parameters are comparable to the research nuclear reactors. Sapphire filter was treated in order to improve the results. Simple and advanced neutron shielding materials considered which was further enhanced with layers of bismuth. The system was compatible with the European Union Directive on ??Restriction of Hazardous Substances?? (RoHS) 2002/95/EC, hence excluding the use of cadmium and lead.     

Trace determination of zirconium using charged particle activation

Reaction yields on thick zirconium targets with proton, deuteron,³He and⁴He activation have been measured and compared. The most favorable reaction combining high sensitivity and selectivity was found to be Zr(p, xn)⁹⁰Nb. The absolute cross-section for this excitation function has been measured for proton energies up to 20 MeV. Analytical applications are illustrated with nondestructive zirconium analyses in glass. Yields from the thick target experiments indicate that the detection limit for this zirconium analysis method is ～1 ppb.     

H*(10) and neutron spectra around linacs

Neutron spectrum and ambient dose equivalent has been measured around two 10 MV linear accelerators. Measurements were carried out in Mevatron and Primus model linacs, both made by Siemens. Main differences between those models are the beam collimator and the vault room. Here, Bonner sphere spectrometer with a passive thermal neutron detector has been utilized to measure the neutron spectrum inside the vault. Using an active detector the neutron spectrum was measured at the vault’s door of both accelerators. With a neutron area monitor the dose equivalent was measured by the door at both vault doors. Neutron strength, total fluence rate and ambient dose equivalent were compared, from this was found that shielding conditions are better in the Primus model.     

A brief review of the determination of cadmium by prompt gamma-ray neutron activation analysis

An examination of the literature concerning the determination of cadmium by Prompt Gamma-Ray Neutron Activation Analysis (PGNAA) has been conducted. In-vivo activation analysis of the liver and kidney is the most common application reported and is briefly reviewed here. This review will concentrate on the determination of cadmium in in-vitro systems. These include a number of different complex matrices such as geological, environmental and biological materials, as well as water, sediments, foods and construction material. Nuclear reactors, accelerators, and radioisotopes have all been used as neutron sources with varying degrees of sensitivity.     

Measurements of fast neutron capture cross sections on <Superscript>63</Superscript>Cu and <Superscript>186</Superscript>W

Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were measured by fast neutron activation method at neutron energies from 1 to 2 MeV. Monoenergetic fast neutrons were produced by ³H(p,n)³He reaction. Neutron energy spread by target thickness, which was assumed to be the main factor of neutron energy spread, was estimated to be 1.5% at neutron energy of 2.077 MeV. Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were calculated by reference comparison method on those of ¹⁹⁷Au(n,γ). Not only statistical errors of gamma-counts from samples but also systematic errors in the counting efficiency for HP Ge detector and the uncertainty of areal density of samples were considered in calculating neutron capture cross section. Estimated neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were also compared with ENDF-6 data.     

Comparative methods for determination of rare earths in geological and mineral samples

Particle induced X-ray emission (PIXE) has been used for the quantitative analysis of rare earth elements (REE) in thick targets prepared from geological and mineral samples. Measurements were made with 1 and 3 MeV proton beams. For comparison, determination of the same elements was made by X-ray fluorescence (XRF) using a²⁴¹Am annular source. Minimal detectable limits (MDL) have been estimated for the two situations. Neutron activation analysis (NAA) has also been used for the determination of REE at the ppm level.     

Multiple fast neutron and gamma-ray beam systems for the detection of illicit materials

Neutron and photon sources have been combined in order to assess the performance of these combinations to discriminate between materials of similar composition. The evaluation has been carried out on the basis of the maximum and minimum ratio values of the relative transmissions of high-energy gamma- or X-ray and neutrons. The number of materials with similar ratio values was used as an indicator of the effectiveness of each source combination. The use of three sources, instead of two, significantly improves the capability of neutron/photon combination to separate similar in composition materials.     

Methods and equipment of industrial activation analysis

Expanding application of activation analysis in industry resulted in exposure of features related to higher requirements to productivity, reliability, automation level, metrological support of analytical methods and equipment. Based on the application of neutron generators, radioisotope neutron sources, nuclear reactors, electron accelerators as activating radiation sources, high-productivity activation analytical systems used directly in analytical laboratories and plants were constructed. Level of development of the above works makes it possible to conclude that industrial activation analysis has formed as an independent trend of nuclear analytics and has considerable prospects.     

Advanced compact accelerator neutron generator technology for active neutron interrogation field work

Due to a need for security screening instruments capable of detecting explosives and nuclear materials there is growing interest in neutron generator systems suitable for field use for applications broadly referred to as active neutron interrogation (ANI). Over the past two years Thermo Electron Corporation has developed a suite of different compact accelerator neutron generator products specifically designed for ANI field work to meet this demand. These systems incorporate hermetically-sealed particle accelerator tubes designed to produce fast neutrons using either the deuterium-deuterium (E _n = 2.5 MeV) or deuterium-tritium (E _n = 14.1 MeV) fusion reactions. Employing next-generation features including advanced sealed-tube accelerator designs, all-digital control electronics and innovative housing configurations these systems are suitable for many different uses. A compact system weighing less than 14 kg (MP 320) with a lifetime exceeding 1000 hours has been developed for portable applications. A system for fixed installations (P 325) has been developed with an operating life exceeding 4500 hours that incorporates specific serviceability features for permanent facilities with difficult-to-access shield blocks. For associated particle imaging (API) investigations a second-generation system (API 120) with an operating life of greater than 1000 hours has been developed for field use in which a high resolution fiber-optic imaging plate is specially configured to take advantage of a neutron point-source spot size of ∼2 mm.     

Bulk material analysis using 14 MeV neutrons

There is considerable interest in bulk material analysis using energetic neutrons e.g. on-line coal analysis and down-hole logging. Recent work has indicated that an in-situ analysis of all the major elements of coal, using prompt capture and inelastic neutron scattering reactions, may be an operational possibility. The application of these techniques to oil well logging would allow the determination of lithology, porosity and oil and water saturation. Gamma-ray spectra arising from 14 MeV neutron bombardment of well characterised coal and fluid saturated rocks are presented. The gamma-ray intensities from both capture and scattering reactions are determined. Neutron transport modelling is used to evaluate the effect of variations in material content and the presence of trace neutron poisons. Predictions of carbon and oxygen scattering response are compared with the experimentally determined ratios.     

Electron beam accelerators—trends in radiation processing technology for industrial and environmental applications in Latin America and the Caribbean

The radiation processing technology for industrial and environmental applications has been developed and used worldwide. In Latin America and the Caribbean and particularly in Brazil there are 24 and 16 industrial electron beam accelerators (EBA) respectively with energy from 200 keV to 10 MeV, operating in private companies and governmental institutions to enhance the physical and chemical properties of materials. However, there are more than 1500 high-current electron beam accelerators in commercial use throughout the world. The major needs and end-use markets for these electron beam (EB) units are R and D, wire and electric cables, heat shrinkable tubes and films, PE foams, tires, components, semiconductors and multilayer packaging films. Nowadays, the emerging opportunities in Latin America and the Caribbean are paints, adhesives and coatings cure in order to eliminate VOCs and for less energy use than thermal process; disinfestations of seeds; and films and multilayer packaging irradiation. For low-energy EBA (from 150 keV to 300 keV). For mid-energy EBA (from 300 keV to 5 MeV), they are flue gas treatment (SO₂ and NO_X removal); composite and nanocomposite materials; biodegradable composites based on biorenewable resources; human tissue sterilization; carbon and silicon carbide fibers irradiation; irradiated grafting ion-exchange membranes for fuel cells application; electrocatalysts nanoparticles production; and natural polymers irradiation and biodegradable blends production. For high-energy EBA (from 5 MeV to 10 MeV), they are sterilization of medical, pharmaceutical and biological products; gemstone enhancement; treatment of industrial and domestic effluents and sludge; preservation and disinfestations of foods and agricultural products; soil disinfestations; lignocellulosic material irradiation as a pretreatment to produce ethanol biofuel; decontamination of pesticide packing; solid residues remediation; organic compounds removal from wastewater; and treatment of effluent from petroleum production units and liquid irradiation process to treat vessel water ballast. On the other hand, there is a growing need of mobile EB facilities for different applications in South America.     

中子俘获治疗癌症*

简要介绍了中子俘获治疗(NCT ) 的原理、发展历史及最新进展, 指出成功地实现NCT 需要高质量的超热中子束、高选择性的NCT 药物、严格的剂量计算和测量及周密的治疗方案。本文对加速器和反应堆产生超热中子束的两种方法及质量要求作了较详细的讨论, 对BNCT、GdNCT、LiNCT、UNCT 及相关的药物进行了比较性评述, 并对我国开展NCT 研究提出了建议。     

Developments in accelerator based boron neutron capture therapy

This paper will review the current status of Boron Neutron Capture Therapy (BNCT), from basic physical mechanisms and clinical indications, to neutron beam development and dosimetry. For in-hospital facilities, particle accelerators presently provide the favoured option, and this paper concentrates on this approach to neutron beam production for BNCT. Various accelerator-based approaches will be reviewed, but discussion will concentrate on the Birmingham programme, particularly the design of a suitable neutron beam delivery system and the experimental validation of Monte Carlo simulations on a mock-up neutron beam moderation system. The use of dose modifying factors to evaluate the likely clinical utility of an epithermal neutron beam will also be discussed, with illustrations from the Birmingham programme.     

Prompt analysis of rare earths by high-energy PIXE

Thin targets of rare earth fluorides were bombarded with 66 and 85 MeV protons. Measured cross sections for X-ray production agreed with PWBA calculations. Satellite X-rays from nuclear reactions were obtained for both (Z-1) and (Z+1) products from the bombardment of element Z. Interference-free sensitivities were of the order of tens of nanograms under bombardment with 1 mC of integrated charge. The technique was applied for the analysis of geological ores and standards.     

IAEA activities for the production and utilization of isotopes: Some examples of recent work

The paper gives a brief overview of typical IAEA activities which contribute to the production of isotopes in nuclear reactors and accelerators, as well as their use. The areas touched upon include (1) isotope production in research reactors and accelerators, (2) quality control and quality assurance in radioanalytical measurements, (3) neutron activation analysis, (4) nuclear methods for land mine detection, (5) radiopharmaceuticals and nuclear medicine, (6) isotope techniques of water resource management and (7) soil management and crop nutrition.     

Determination of oxygen in aluminium by means of 14 MeV neutrons with an account of flux attenuation in the sample

This study concerns the determination of oxygen present at low levels in aluminium using the 14 MeV neutron activation technique. Neutron attenuation effects in the aluminium samples comprising scattering and slowing down processes below threshold energy have been treated.     

Computational study of pulsed neutron induced activation analysis of cargo

Pulsed neutron induced activation analysis is a nondestructive technique to detect threats hidden in bulk objects such as cargo pallets, trucks, etc. Isotopic content of cargo can be measured by counting photons emitted with characteristic energies as a result of neutron induced reactions within cargo’s materials. Neutron and gamma radiation transport in active interrogation system consisting of a 14-MeV neutron source, photon detector, and a cargo truck was analyzed with MCNPX code. Gamma ray signatures of cargo with hidden explosive threat were analyzed during the neutron pulse and between neutron pulses for varying system’s geometry and material composition of cargo.