Metallic Composition Analysis of Crude Petroleum from Some Oil Fields in Ghana

Qualitative and quantitative analyses of chemical elements in crude petroleum using energy-dispersive X-ray fluorescence spectroscopic technique has attracted the attention of scientific world because it is fast, cheap, non-destructive and assurance in quality compared to other methods. Metallic element characterisation of crude petroleum is important in the petrochemical industry because it determines rock reservoir properties, the technology needed for extraction and refinery process, hence an exciting field that calls for research. X-ray fluorescence method was used for metallic composition analysis of four rundown crude petroleum samples (SB-2, SB-4, TB-2 and TB-1) from three oil fields (Saltpond, TEN and Jubilee). It was conducted at the National Nuclear Research Institute of Ghana. Analysis of the four samples concluded that oil field maturity decreases orderly from Saltpond, Jubilee and TEN. Vanadium-nickel ratios for each crude petroleum sample was less than 0.5, indicating that both Saltpond and Tano sedimentary rocks are of marine organic origin. Higher concentration levels of rare earth metal elements (scandium and yttrium) in the Saltpond sedimentary basin compared to Tano sedimentary rock suggest seismic effect of McCarthy Hills on Saltpond Basin. The strong negative correlation between the vanadium-nickel ratio (predictor) and scandium concentration (dependent) among the three oil fields implies that scandium concentration can equally be used to characterise the oil fields just as the vanadium-nickel ratios.

Using a delayed coincidence counting system to determine ²²³Ra, ²²⁴Ra in seawater sample

A Radium Delayed Coincidence Counter (RaDeCC) includes 3 channels (223Ra channel,224Ra channel, and total channel). It has been newly designed and assembled at Nuclear Research Institute. To determine 223Ra and 224Ra in seawater samples, the system efficiency at all 3 channels were investigated and calibrated. The research results showed that the RaDeCC operates stably and reliably with high efficiency of 26%. In this project, a procedure for measuring short half-life radium isotopes was established with a low detection limit (LOD (223Ra) = 0.002 Bq; LOD (224Ra) = 0.01 Bq), good reproducibility, and high precision. The technique is suitable for qualitative analysis of 223Ra, 224Ra in seawater samples at low concentration. The 11 coastal water samples were collected in a coastal of Ninh Thuan province. The analytical data of short-lived radium isotopes concentration in seawater at Ninh Thuan coastal area are 11.2 × 10-3 ÷ 45.5 × 10-3 mBq/L for 223Ra, and 34.7 × 10-2 ÷ 21.9 × 10-1 mBq/L for 224Ra.

Dose Calculation and Measurement from B¹⁰(n, α)Li⁷ Reaction Using Filtered Neutron Beam at Nuclear Research Institute

In this research, dose calculation and measurement from B10 (n, α) Li7 reaction usingfiltered neutron beam at the Nuclear Research Institute have been reported. Calculation was carried out by Monte Carlo method using MCNP5 code. Neutron activation technique using vanadium foil was employed to determine neutron flux at various positions in phantom from which neutron dose has been calculated using conversion factor. These calculations are basics for the dose determination research of the Boron Neutron Capture Therapy (BNCT) in Vietnam.

Quantification of effluents in the production of nuclear fuel

The Brazil with the purpose of becoming self-sufficient in the production of radioisotopes and radioactive sources used in nuclear medicine, agriculture and the environment developed the project of a multipurpose reactor of 30 megawatts of power to meet the national demand. In the Energy and nuclear research Institute (IPEN), Nuclear fuel Center (CCN) is responsible for the manufacture of fuel for the reactor IEA-R1 and, possibly, by the multipurpose reactor fuels. In order to meet the demand for the reactors was designed a new manufacturing plant with a maximum capacity of 60 fuel per year, which currently is 10. The increase in production as a consequence will increase the volume of effluents generated. The current concern with the environment it is necessary to draw up a management plan to make the process sustainable, which will result in environmental, economic and social benefits. The fuel production process generates various types of effluent containing uranium or not – being solid, gaseous and liquid with different physical and chemical characteristics. The objective of this work is to follow the process of nuclear fuel production and to identify, quantify and characterize the effluents, especially liquids, to subsequently draw up a management plan and eventually dispose of responsibly in the environment.

VVER-1000 BENCHMARK INTERPRETATION WITH MONTE CARLO CODE MCS

An interpretation of the NEA-1517/82 benchmark from the SINBAD shielding database has been conducted with the MCS Monte Carlo code developed at the Ulsan National Institute of Science and Technology (UNIST) and the ENDF/B-VII.1 nuclear data library. The NEA-1517/82 benchmark corresponds to experiments on a VVER-1000 critical mock-up (thermal reactor with hexagonal fuel lattice) inside the LR-0 research reactor operated by the Nuclear Research Institute (NRI) in the Czech Republic. A new 3D model of the VVER-1000 mock-up core is developed for MCS based on the SINBAD documentation. The model includes the top and bottom parts of fuel pins, the spacer grids and core components: baffle, barrel, downcomer, tank, reactor pressure vessel (RPV) and concrete block used as biological shielding. The quality of the model is verified first by code/code comparison of MCS against MCNP6 for criticality and power distributions (pin-by-pin and axial power). The validation of MCS results is then performed against six critical cases, 260 measured pin powers and benchmark calculations of the axial power profile. Finally, a comparison of calculated and measured neutron spectra inside the mock-up core is presented as a preliminary study for upcoming works on the deep-penetration shielding capability of MCS.

An algorithms to improve the energy resolution of two-step cascade spectrum

The present paper proposes an algorithm to improve the energy resolution of two-step cascade spectrum. The energy resolution plays an important role in the domain of gamma spectrum analysis. The better the energy resolution is, the better the ability of peak resolving is. The algorithm is constructed based on an analyze of energy resolution of the summation amplitude of coincident pulses spectrometer using the analogue technique. The algorithm proposed has been tested on some two-step cascade spectra of 164Dy nucleus obtained from the (n, ) reaction experiment using the gamma – gamma coincidence spectrometer at Dalat Nuclear Research Institute. Two-step cascade spectra corresponding to the cascade decays from the compound state to final states whose energies are 0, 74, and 242 keV have been evaluated. The results obtained show that the energy resolution of the two-step cascade spectrum has been reduced by 1.05 to 2.04 times within the energy range of 586 to 6830 keV. Our algorithm can therefore be applied to improve the ability of peak deconvolution, the accuracy, and the realibility in analyzing two-step cascade spectra.

Estimating partial body ionizing radiation exposure by automated cytogenetic biodosimetry

PurposeInhomogeneous exposures to ionizing radiation can be detected and quantified with the Dicentric Chromosome Assay (DCA) of metaphase cells. Complete automation of interpretation of the DCA for whole body irradiation has significantly improved throughput without compromising accuracy, however low levels of residual false positive dicentric chromosomes (DCs) have confounded its application for partial body exposure determination.Materials and MethodsWe describe a method of estimating and correcting for false positive DCs in digitally processed images of metaphase cells. Nearly all DCs detected in unirradiated calibration samples are introduced by digital image processing. DC frequencies of irradiated calibration samples and those exposed to unknown radiation levels are corrected subtracting this false positive fraction from each. In partial body exposures, the fraction of cells exposed, and radiation dose can be quantified after applying this modification of the contaminated Poisson method.ResultsDose estimates of three partially irradiated samples diverged 0.2 to 2.5 Gy from physical doses and irradiated cell fractions deviated by 2.3-15.8% from the known levels. Synthetic partial body samples comprised of unirradiated and 3 Gy samples from 4 laboratories were correctly discriminated as inhomogeneous by multiple criteria. Root mean squared errors of these dose estimates ranged from 0.52 to 1.14 Gy2 and from 8.1 to 33.3%2 for the fraction of cells irradiated.ConclusionsAutomated DCA can differentiate whole-from partial-body radiation exposures and provides timely quantification of estimated whole-body equivalent dose.Biographical NoteBen Shirley M.Sc. is Chief Software Architect, CytoGnomix Inc. Canada; Joan Knoll Ph.D. Dipl.ABMGG, FCCMG is Professor in Pathology and Laboratory Medicine, Schulich School of Medicine and Dentistry, University of Western Ontario, Canada and cofounder, CytoGnomix Inc.; Jayne Moquet Ph.D. is Principal Radiation Protection Scientist in the Cytogenetics Group, Public Health England; Elizabeth Ainsbury Ph.D. is Head, Cytogenetics Group and the Chromosome Dosimetry Service, Public Health England; Pham Ngoc Duy M.Sc. is deputy director of Biotechnology Center, Dalat Nuclear Research Institute, Vietnam; Farrah Norton M.Sc.is Research Scientist and Lead of the Biodosimetry emergency response and research capability at Canadian Nuclear Laboratories; Ruth Wilkins, Ph.D. is Research Scientist and Chief of the Ionizing Radiation Health Sciences Division at Health Canada, Ontario, Canada; and Peter K. Rogan Ph.D. is Professor of Biochemistry and Oncology, Schulich School of Medicine and Dentistry, University of Western Ontario, Canada, and President, CytoGnomix Inc.

Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.