Event Schedules

  • Day 01

    Apr 16, 2019

  • Day 02

    Apr 17, 2019

Keynote Speakers

Carmen Tuca

University of Craiova

Preparatory steps for VVR-S Nuclear Research Reactor Hot Cells dismantling

The paper presents steps performed prior to the dismantling of the Hot Cells from IFIN-HH VVR-S Nuclear Research Reactor. The decommissioning is the final step in the lifecycle of the nuclear facility after the operating completion and final shutdown. Reactor was operated between 1957 and 1997 at a nominal thermal power of 2MW. The main purpose was the radioisotope production for industrial and medicine applications also the research activities in physics, biophysics and biochemistry. The reactor decommissioning started in 2010 and will be completed in 2020. The main steps in the Hot Cells decommissioning consists of the clean-up (radioactive waste evacuation), radiological characterisation of the internal parts (e.g. the fixed equipment and stainless-steel lining), surfaces decontamination, internal parts dismantling and evacuation. The dismantling strategy was chosed according to the radioactive inventory of the released waste also the radiological characterisation results. Before decommissioning the inventory was estimated at about 14.9 Ci. The main contaminants (with a half-life longer than one year) that contribute significantly to the inventory are: 60Co, 134Cs, 137Cs, 152Eu and 108mAg. After the evacuation the radioactive inventory is 10.6 Ci. The radiological characterisation was the most important task implemented prior to the hot cells dismantling and consisted of: radiation background and equivalent dose rate measurements; contamination measurements to determine the surface activity of the objects/internal spaces such as alpha-beta scanning and pair measurements; in-situ gamma spectrometric analysis for radionuclide composition identifying and smear sampling analysis using a spectrometric system with NAI(TL) detector.

Ali Alghamdi

Immam Abdulrhman Bin faisal University

Nanoparticles for enhancing dose delivery ;Simulation and Experimental study using Cancer Cells

Gold Nano Particles (GNP) have many potential uses for cancer therapy. In particular interest the physical characteristics of Gold i.e. higher density (Z) and electron conductivity. Once synthesize in the form of Nano scale particles they pose higher surface areas. The use of GNPs in this study is investigated at single cell model and tumor model using MCNPX 2.5.0 Monte Carlo code. In order to increase the probability of X-ray interaction in the cell vicinity a force collision card is used in the MCNPX file. The cell geometrical cards important in the MCNPX file were adjusted to allow more interactions of X-ray photon within the tumor cell cytoplasm and within GNPs. The same model used to simulate thermal neutron beam with boron nano particles embedded in the tumor cells. Energy deposition were tallied in a single cell using a very fine mesh tally of 3D energy bins coordinate in 13 micrometer cell diameter and in the 13cm tumor comprising 4 million tumor cells. In this study the experimental part applied by using cancer cells with and without GNP as they were exposed to X ray using the normal medical diagnostic. Some modifications were employed to enhance the dose by removing the beam filter. Tow ranges of energies were used and the cell counting took place after 48 hr from the exposure. The doses were estimated based on the exposure factors of 80Kvp and 40 Kvp with a fixed 320MAS delivering an estimated dose of 2.9cGray per exposure. Minimum possible distance between the X-ray tube and the cultured cell dishes was achieved by removing the collimation box. Results of the simulations and experiments shows the feasibilities of gold and boron nano particles for enhancing dose to the tumor cells.

Hirose K

Department of Materials and Life Sciences, Faculty of Science and Technology, Sophia University, Japan

Fukushima nuclear accident: New aspects of environmental radioactivity monitoring and nuclear techniques

Fukushima Daiichi Nuclear Power Plant (FDNPP) accident, which occurred in March 2011 due to The Great East Japan Earthquake and resulting gigantic tsunami, has contaminated atmosphere, terrestrial and marine environments by enormous releases of radioactive materials. After the FDNPP accident, radioactivity in environment has been measured by using various methodologies. Information on levels of radiation, their spatial distributions, and concentrations of radionuclides in environmental samples is important to implement radiation protection and to elucidate environmental behaviors and fate of FDNPP-derived radionuclides. The aerial measurement has provided important information on the spatial distributions of radiation dose and major deposited radionuclides (134Cs and 137Cs) not only in land area but also in sea area. The traditional monitoring methodologies such as radiometric methods have been deployed radioactivity measurements of environmental samples, in which short and long-lived gamma emitters (131I, 132I, 134Cs, 137Cs, 110mAg and others) in atmosphere, seawater, soil and biota have been measured by gamma spectrometry. Mass spectrometric techniques such as ICP-MS, AMS and others have been applied measurements of small amounts of the long-lived fission products (129I, 135Cs) and fissile materials (uranium, plutonium, americium and curium isotopes) in addition to alpha spectrometry. Chemical speciation of radionuclides released in the environment from the FDNPP is important to have better understanding of their behaviors in environment. Hot particles as did the Chernobyl accident were detected in air and soil samples, which were studied by scanning electron microscopy with energy dispersive X-ray analysis. The results reveal that the environmental impacts of the FDNPP accident continued over 7 years, although remediation of damaged reactors and decontamination of radionuclides in soil have been implemented extensively.

Petr G. Sennikov

G.G.Devyatykh Institute of Chemistry of High-Purity Substances of RAS, Russian Federation

Chemical and plasma-chemical preparation of stable isotopes of molybdenum, boron and silicon for nuclear medicine

Stable isotopes are used usually in nuclear medicine as precursors of radioactive isotopes that serve straight as medicine or imaging agents. Well known examples of these isotopes are 98Мо и 100Мо (concentration in natural Mo is equal to 23,75 и 9,62%, correspondingly) that form after irradiation with neutrons or protons first radioactive isotope 99Mo and then isotope 99mTc serving as imaging agent in cardiology and oncology. Another example is 10В (19,9%) isotope that after irradiation with slow neutrons forms first radioactive 11В and than stable 7Li together with alpha-particles killing the tumor cells. From the other side, the rare stable 29Si (4,71%) isotope is perspective per se as delivery agent of drugs and as magnetic resonance imaging (MRI) substance for detecting of cancer tumors. In this lecture the new method of manufacturing of above mentioned stable isotopes as well as others having volatile fluorides will be discussed. The last condition is important because of high efficiency of centrifugal method of isotopic separation using fluoride molecules (fluorine is monoisotopic). The conversion of enriched fluorides to target isotopes can be carried out either chemically via intermediate substances or in one step using plasma-assisted approach. Some preliminary results concerning neutron irradiation of 98Mo nano-particles for production of 99Мо and application of 29Si nano-particles as contrast agent for MRI of mice with cancer tumors in vivo will be discussed.

Sessions:

Nuclear Chemistry,Radiation Chemistry, Nuclear Engineering, Nuclear Reactor, Nuclear Fusion, Nuclear power, Nuclear Chain Reactions,Nuclear Fuel cycle

Session Chair:

1

Nicole Virgili

Sapienza University, Rome,Italy

VERIFYING THE PERFORMANCE OF THE ARGON-41 MONITORING SYSTEM FROM FLUORIDE-18 PRODUCTION FOR MEDICAL APPLICATIONS

In this work, the well-known MC code was used to simulate the TR19PETcyclotron (19 MeV) installed at “A. Gemelli” University Hospital (Roma, IT) and routinely used in the production of positron emitting radionuclides. In a medical cyclotron facility, 41Ar (t1/2 = 109.34 m) is produced by the activation of air due to the neutron flux during irradiation, according to the 40Ar(n,γ)41Ar reaction; this is particularly relevant in widely diffused high beam current cyclotrons for the production of PET (Positron Emission Tomography) radionuclides, 18F radionuclide in this case. The aim of this work is the determination of the detection efficiency of a Geiger Muller detector placed in the terminal part of the chimney of the cyclotron for environmental monitoring of 41Ar emission through the chimney of the cyclotron. Function of the detector is to reveal the activity of 41Ar produced by the cyclotron. Taking into account the activity of 41Ar inside the bunker of 700 Bq/μA, beam current intensity of 50 μA and 41Ar radioactive decay constant of 1,054 · 10-4 s-1, the activity of 41Ar emitted and the detection efficiency have been calculated. In conclusion the detection efficiency determinated through the use of Monte Carlo code is very low (4,7 10-7 cps/Bq), consequently the counting rate of 1,3 · 10-2 is low and several studies have now been conducted to find more effective detectors.

Boris Gurovich

NRC "Kurchatov Institute", Moscow, Russian Federation

At present, the lifetime of third-generation nuclear power reactors (VVER-1000) is about 60 years. For reactors Generation III+ this lifetime is expected to extend from 80 to 100 years. Since the reactor vessel is an irremovable element of the reactor installation, the lifetime of the reactor materials determines the operation possibility of the NPP. In this work, complex microstructural studies and mechanical tests of new steels compositions (with a reduced nickel concentration) of VVER reactors in various states were carried out: initial, after provoking embrittlement heat treatment in the interval of temper brittleness maximum (formation of grain-boundary segregations), as well as after accelerated neutron irradiation up to fluences 100•1022m-2. It was shown that these steels are distinguished by higher thermal stability and radiation resistance at operating temperature as compared with VVER-1000 base metal, what allowed justifying their safe operation for 80-100 years.

Alex Frolov

NRC "Kurchatov Institute", Moscow, Russian Federation

Radiation-induced features of the zirconium shells of fuel elements

In a number of works, it has been shown that during neutron irradiation in the materials based on Zr the evolution of the structural-phase state is observed, with the formation of second phases and radiation-induced dislocation loops, which can form periodic structures. This paper presents the results of microstructural studies of irradiated fuel element claddings of the E110 alloy (with the addition of 1% niobium) after operation in the fuel assemblage (burnup ~28-44 GWd/tU). This causes the degradation of mechanical properties - plasticity reducing, yield strength increasing, etc. To understand the mechanisms of properties degradation of E110 fuel element under the influence of reactor irradiation, the complex microstructural studies are necessary. The following structural elements were found in the studied samples: β-Nb particles, particles based in niobium and zirconium (Nb-Zr), Laves cubic phases based (Zr(Nb,Fe,Cr)2), δ-hydrides, γ-hydrides, -type dislocation loops, as well as dislocation structures with -component. Finely dispersed phases (coherent with the matrix) based on niobium, having a needle-like shape, were also found, and which are formed a two-dimensional periodic structure of precipitates aligned along the basal planes of HCP Zr lattice. It is established that at the initial operational stage these phases have an HCP lattice with parameters close to the α-Zr lattice, with an increase of irradiation dose the lattice transforms into a bcc type, which is typical for β-Nb. Dose dependences of the densities and sizes of the phase components were also obtained.

Evgenia Kuleshova

NRC "Kurchatov Institute", Moscow, Russian Federation

Influence of the irradiation temperature on structural features of pressure vessel steels

A set of structural studies and mechanical testing of VVER-1000 steel in the initial state, after accelerated irradiation at temperatures of 50°C, 150°C, 300°C and 400°C, and also after provoking embrittlement heat treatment was carried out. The phase composition and mechanisms of the radiation embrittlement were estimated for the studied steel after accelerated irradiation in the temperature range of (50-400)°C and fast neutron fluences (5.1-45.3)∙1022 m-2. It was shown that in the temperature range of (50-140)°С, there is no radiation-induced phases and grain-boundary segregations, but there is the formation of radiation defects (dislocation loops) with a quite high number density. The formation of dislocation loops and radiation-induced precipitates is observed at the operating temperature of the WWER-1000 (300°C), as well as the accumulation of thermally and radiation-stimulated grain-boundary segregation of phosphorus. It was shown that accelerated irradiation at 400° C leads to the absence of radiation defects and radiation-induced precipitates, as well as to an increase of the phosphorus concentration on grain-boundaries and carbonitrides density. Mechanical tests showed that the steel has the maximum shift of ductile-to-brittle temperature after irradiation at 400°C, and the maximum change in the yield strength is observed after irradiation at 50°C.

Grzegorz Zuzel                

Jagiellonian University, Poland

Determination of Po-210 in materials down to single mBq/kg.

Surface and bulk contamination with long-lived daughters of Rn-222 is of great interest for most of the experiments looking for rare events. These include the detection of low energy solar neutrinos in real time, searches for neutrino-less double beta decay or searches or dark matter. Decays of Pb-210, Bi-210 and finally Po-210 may contribute significantly to the experiments’ background, especially when they appear close (external background coming from e.g. construction materials/shields) or directly in the active volumes (internal contamination). Two methods to determine the bulk Po-210 will be presented. The first is based on direct counting with a large surface ultra-low background alpha spectrometer. By attributing the counts in the registered spectra in the range of 1.5 MeV to 6 MeV to sub-surface Po-210 and by application of the Monte Carlo simulations we are able to investigate the bulk Po-210 contamination (alphas coming from different depths can populate the spectrum up to the 5.3 MeV, which including the energy resolution of the device was extended to 6 MeV) down to 50 mBq/kg. The second method is based on chemical separation of Po form the bulk material on a dedicated column and subsequent counting of Po-210 (deposited on a silver disc) with a dedicated low-background spectrometer. First test show that the sensitivity, which may be achieved here after optimization of the procedure, is at the level of single mBq/kg. Thus, this would be presently the most sensitive method to determine the Po-210 specific activities. Measurement of Po-210 content for various materials, like copper of different chemical purities, titanium or steel, will be presented. By investigating the same material in a time sequence it is also possible to establish the equilibrium conditions in the Pb-210 – Po-210 U-238 Sub-chain.

Abdel-Mjid Nourreddine

Institut Pluridisciplinaire Hubert Curien, UMR

Beta Energy Dependence Response Of The Ag-Doped Phosphate Glass Detectors

The radiophotoluminescent signal of Ag-doped phosphate glass detectors to beta radiation is significantly influenced by the energy. As consequence, the detector does have not the same sensitivity for the same absorbed dose, especially for the reference beta energies specified by ISO 6980-1. An accurate dose assessment could be biased if the calibration was performed with beta rays other than a beta field of the monitored area. In this work we propose to investigate the beta energy dependence of the emitted radiophotoluminescence in Ag+-doped phosphate glass detectors. The results show that the glass sensitivity decreases with energy. The angular relative error to 90Sr/90Y in comparison to the reference beta particle radiation was calculated using the MCNPX code. Good agreement was observed between the experimental data and calculated values. A relative error going up to more 70% was observed.

Sessions:

Nuclear Materials ,Nuclear Medicine, Nuclear Safety,Radioactivity ,Radiatio Monitoring,Radiobiology,Radio Pharmacology,Radiation Protection

Session Chair:

2

P. Nerisson

Institut de Radioprotection et de Sûreté nucléaire

VOLATILIZATION AND TRAPPING OF RUTHENIUM UNDER A L

The reprocessing of spent nuclear fuel produces high level liquid waste (HLLW). Due to the decay heat, these concentrated nitric solutions containing fission products are stored in cooled tanks to prevent the solution from boiling, evaporating and drying out, which could lead, in case of loss of cooling, to large releases of radioactive materials into the environment, especially ruthenium volatile species. In the post-Fukushima complementary safety assessments, the loss-of-cooling accident on HLLW storage tanks is one of the accident scenarios identified as a dreaded situation. It is also taken into account in defining the on-site emergency plan in Orano La Hague (France) reprocessing plant. Besides, an extensive literature review performed at IRSN confirms the lack of reliable data on the behaviour of ruthenium in nitric acid solutions. It highlights that research works on this topic can be classified in several categories: ruthenium chemistry in a nitric medium characterized by the formation of nitrosyl ruthenium ion RuNO3+; behaviour of volatile forms of ruthenium in presence of steam, nitric acid and nitrogen oxides (recombination, decomposition, etc.); transfer phenomena of the different gaseous species containing ruthenium. Subsequently, the efficiency and the performance of various systems that can be used to trap and filter ruthenium (gas/liquid absorbers, steel filters, zeolites, etc.), or even prevent its volatilization (recombination, addition of reducing agents in situ, etc.) have been investigated by different authors. Previous work performed at IRSN on severe accident scenarios in the nuclear power plant context allowed to characterize the filtration devices existing in the containment venting line of French PWRs, i.e. the sand filter and the metallic prefilter, with respect to RuO4(g). It showed that these latter do not trap efficiently RuO4. From these findings, IRSN started a research program aiming at improving the knowledge on this topic. A specific test bench has been developed in order to study the volatilization of a nitric acid solution containing Ru nitrosyl, simulating a real HLLW in terms of acidity and ruthenium concentration, and to investigate the possible inhibition of Ru volatilization by addition of specific reducing compounds. The experimental device is also used to study RuO4 trapping by different materials: zeolites, MOFs (Metal Organic Framework) and active charcoals. First experimental results are presented.

Muhammad Saeed

the effect of injecting air-fountain inside a containment to investigate the gas distribution analysis by using HYDRAGON code

This paper presents the effect of injecting air-fountain inside a containment to investigate the gas distribution analysis by using HYDRAGON code. The effect of three turbulence models, i.e., a standard k − ε model, a Re-Normalized Group (RNG) k − ε model and a realizable k − ε model was analyzed and the simulation results were compared with the published experimental data. Three different air-injection velocities were used to analyze the stratification break-up phenomena.When the air-injection velocity was set to 0.411 m/s and 5.143 m/s, the simulation results obtained by using all the three turbulence models were in better agreement with the experimental data.However, when the air-injection velocity was set to 2.803 m/s, the simulation results obtained by SKE and RNG turbulence models have captured the experimental trends at different elevations better than RLZ turbulence model. When the turbulent diffusivity coefficient was applied to RNG and RLZ turbulence models, a small effect on simulation results was observed. Moreover, the results obtained by using SKE turbulence model with turbulent diffusivity coefficient term had no noticeable effect.

Noeen M

TRIUMF & BCCRC, Canada

Advances in Radiochemistry & Clinical Applications of PET Radionuclides

Among the earliest radiotracers developed for PET, [18F]FDG stands as a striking example which reflects the significance of 3D-imaging for clinical purposes. Its availability largely depends either on onsite production or delivery from nearby. With the passage of time and advances in radiopharmaceutical production, several more radiotracers such as [18F]FDOPA, [18F]FLT, [11C]Choline, [11C]PIB, [68Ga]DOTA-TATE and [68Ga]PSMA-HBED-CC have been in clinical use since more than a decade. The most recent interest in [89Zr]-labeled antibodies technology has a wide impact in cancer immunology research. To target a wide range of patients, production of PET radiotracers via automation plays a key element as it enables a feasible access to the required doses. For facile cGMP production, new innovative modules such as FASTlabs have paved the way to scan scores of patients per day. Contrary to [18F]-, [11C]- and [89Zr]- production via onsite cyclotron, the dependence of [68Ga]- production on availability of [68Ge/68Ga] generators, has limited its large-scale utility. However, with the ongoing development of [Al18F]- labelling strategy, several tracers such as PSMA-HBED (with [Al18F]-) are in the early phase of clinical research.

Radioactivity in the Oil Exploration Sector

Sheldon Landsberger

University of Texas at Austin, Nuclear Engineering Teaching Lab, Austin, Texas, USA, 78757

One of the very first papers describing radioactivity in oil extraction appeared in 1906 just a scant eight years after its discovery by Henri Bequerel in 1896. The world currently consumes closed to 100 million barrels of oil daily and is produced in countries throughout the globe through onshore drilling which refers to drilling deep holes under the earth's surface and offshore drilling which relates to drilling underneath the seabed. It was only in the late 1970’s and early 1980’s where a significant amount of research was done in characterizing the radioactivity in extraction processes which included, scale, produced water, sludge, etc. What is more surprising than the unexpected amounts of radioactivity in the oil extraction sector is the orders of magnitude differences of radiation from different onshore fields. Thus handling of these radioactive by products including transportation, clean-up procedures, and burial requires stringent training and monitoring procedures. For instance, typical dosimeters that are placed on chests while most of the radiation emanates from the ground or lower parts of sludge tanks underestimates the dose to the body. Our previous MCNP calculations have confirmed this assumption. A detailed overview of radiation protection guidelines for the oil exploration sector including analytical measurements of the by-products will be presented.

Nuraddeen Nasiru Garba

Department of Physics, Ahmadu Bello University, Zaria, Nigeria

Radiological Map Of Northern Zamfara State, Nigeria

This study produced the radiological map of Northern Zamfara State, Nigeria using the measured gamma radiation dose rates in the area. Measurements were carried out using a portable survey meter, Inspector Alert Nuclear Radiation Monitor and the coordinates of each point was recorded using a global positioning system. A total number of 166 average measurements were taken with the survey meter held at 1 m above the ground, with at least five readings taken at each location in order to minimize error. The mean TGR rate of the study area was found to be 32 nGy h-1 which is less than the world average value of 59 nGy h-1. Kaura Namoda local government area has the highest mean value of TGR rate of 38 nGy h-1, while Bakura local government area has the lowest mean value of TGR rate of 28 nGy h-1. Radiological map of the study area which shows the distribution of the TGR throughout the area was plotted. Radiological health hazards; outdoor annual effective dose rate, mean population weighted dose rate, annual collective effective dose, lifetime dose and excess lifetime cancer risk were computed as 0.04 mSv y-1, 32.45 nGy h-1, 59.33 Sv y-1, 2.75 mSv, and 1.38 × 10-4 respectively.

FTIR Study and UV-VIS Spectroscopy of nuclear trac

Doaa H. Shabaan

Physics Department, Faculty of Women for Art, Science and Education, Ain Shams University, Cairo, Egypt

The purpose of this work was obtaining information about the interaction of γ- ray with CR-39 track detector by using the UV-Vis spectrometry and FTIR which can be used in concerning sensor for gamma irradiation. CR-39 samples irradiated by 60Co source at different doses. The UV-Vis spectroscopy show transitions electronic in the visible region from ground state to excited state, by increasing gamma doses the absorbance spectrum for all samples take the same behavior with slightly shift. This shift due to decrease in the optical band gap energy Eg. The FTIR spectra show for all samples by number of peaks at 1817, 2367, 2645 and 3234 cm-1 is belong to carbonate group C=O stretching vibration, O=C=O asymmetric stretching, C-H Stretching, H2O free stretching vibration, respectively, by increase gamma ray did not observed changes in the CR-39 groups but observed change in the intensities of peaks.

Yet to be updated