Showing papers in "World Journal of Nuclear Science and Technology in 2017"

Quality Assessment Statistic Evaluation of X-Ray Fluorescence via NIST and IAEA Standard Reference Materials

Abstract: The aim of this study is a quality assessment of X-ray fluorescence laboratory located at the University of Khartoum. The X-ray fluorescence spectrometer system consists, a set of three 109Cd sources of an initial nominal activity of 10 μCi, and Si(Li) detector Energy Dispersive XRF(EDXRF) systems. It is important to carry out this work because it has an effective contribution for a wide range of research and services. The assessment was carried out by measuring 8 NIST-2709a (soil) and 13 IAEA-155 (milk powder) standard reference material samples for repeatability examinations to test the measurement precision. The total combined standards uncertainty values for XRF lab were estimated by an error from repeatability measurements adding 2.6% for error propagation related to the method. For accuracy assessment, three standard statistic approaches were applied, i.e. the Bias %, zeta-score, and En-number. The bias of all elements for both standard materials was found to be within a deviation range from −28% to 7.8%. The results of all elements for both the zeta-score test and En-number have satisfactory results except Th (Thorium) and Zr (Zirconium) which consider as questionable results for NIST SRM 2709a and unsatisfactory results for En-number.

Evaluation of Corrosion Product Activity in a Typical PWR with Extended Cycles and Flow Rate Perturbations

Abstract: Effect of flow rate perturbations has been studied using the modified computer program CPAIR-P for time dependent corrosion rates in Pressurized Water Reactors (PWRs) having extended cycles. In these simulations, a decrease in the corresponding saturation values of corrosion product activity (CPA) is observed for higher pH values. Comparison of CPA’s behavior has been done for constant flow-rate case as well as for transients with elevated 10B levels (~40%) in dissolved boric acid in coolant in two operating cycles. When the flow rate is decreased in the first cycle, the saturation value of CPA attains new higher values. Also, in the second operating cycle, the saturation values are about 12% higher when compared with the values in the first cycle.

The role of nuclear power from a system engineering standpoint

Abstract: This work proposed the application of system engineering methods to identify organizations vital for society, seeking development and well-being. System engineering requires the identification of blocks (or systems), identification of their service functions, identification of states, identification of required quality and identification of constraints. Analyzing modern societies, vital functions were identified and countries behavior was modeled, identifying their states. In this context, nuclear power was found to be fundamental for development and defense because of its inherent advantages for military naval purposes at war. Another striking conclusion is that nuclear power is the best solution for country energy security, more than to avoid climate changes, but to help the nation to resist climate changes. A solution to mitigate the high overnight costs of nuclear power was also proposed. It was demonstrated qualitatively that the adoption of dual purpose mobile nuclear power plants military performances, economic development and risks management.

Development of Diagnostics for High-Temperature High-Pressure Liquid Pb-16Li Applications

Abstract: Liquid lead-lithium (Pb-16Li) is of primary interest as one of the candidate materials for tritium breeder, neutron multiplier and coolant fluid in liquid metal blanket concepts relevant to fusion power plants. For an effective and reliable operation of such high temperature liquid metal systems, monitoring and control of critical process parameters is essential. However, limited operational experience coupled with high temperature operating conditions and corrosive nature of Pb-16Li severely limited application of commercially available diagnostic tools. This paper illustrates indigenous calibration test facility designs and experimental methods used to develop non-contact configuration level diagnostics using pulse radar level sensor, wetted configuration pressure diagnostics using diaphragm seal type pressure sensor and bulk temperature diagnostics with temperature profiling for high temperature, high pressure liquid Pb and Pb-16Li applications. Calibration check of these sensors was performed using analytical methods, at temperature between 380°C - 400°C and pressure upto 1 MPa (g). Reliability and performance validation were achieved through long duration testing of sensors in liquid Pb and liquid Pb-16Li environment for over 1000 hour. Estimated deviation for radar level sensor lies within [−3.36 mm, +13.64 mm] and the estimated error for pressure sensor lies within 1.1% of calibrated span over the entire test duration. Results obtained and critical observations from these tests are presented in this paper.

Response of COX2/PGE2 Inflammatory Pathway to Brown Seaweed Extract in Rats Exposed to Gamma Radiation

Abstract: Background: Systemic inflammation due to radiation exposure has been identified in a biological system by certain metabolic and behavioral disorders. These anarchies mostly mediated under a regulation of cyclooxygenase 2 (COX2) induced production of an inflammatory mediator prostaglandin E2 (PGE2). Aim: This study was undertaken to investigate the anti-inflammatory impact of brown sea weed extract (BSWE) against induction of COX2/PGE2 inflammatory pathway in gamma-irradiated rats. Rats were orally administrated with BSWE (27 mg/kg body weight/day) for 7 consecutive days before exposure to 8 Gy fractionated gamma radiation (2 Gy × 4; every 3 days). Treatment with BSWE was extended along with and in-between irradiation doses for another 14 successive days. Our data demonstrated that the administration of BSWE to rats exposed to gamma radiation, following the regimen suggested, significantly neutralize the changes induced in the inflammatory molecules COX2, PGE2, tumor necrosis alpha (TNF-α), and nitric oxide (NO). In addition, it adjusted significantly the cellular redox tone via regulation of changes induced in malondialdehyde (MDA) reduced glutathione (GSH), superoxide dismutase (SOD) catalase (CAT) and xanthine oxidoreductase system (XOR). Credibly, from the results emerged in this study, it could be suggested that BSWE has substantial anti-inflammatory activities and gamma radiation protection capabilities. It is recommended to include BSWE in the treatment strategy of various inflammatory diseases especially cancer as a safe natural anti-inflammatory agent.

Control Rod Calibration and Worth Calculation for Optimized Power Reactor 1000 (OPR-1000) Using Core Simulator OPR1000

Abstract: Control rod is used to change the power in nuclear reactor. Certainly, the core at any moment can be made subcritical condition and shut downs when occurring to emergency instance in the core. The rod is grouped based on their function and located at different places in the core where their feature is maximized. Two methods of control rod calibration are the asymptotic period method and the rod-drop method, which were applied in this experiment. In the first method, the reactor is made supcritical by inserting the control rod to be calibrated a certain level. The rod drop method is to determine the subcritical; at the critical state, the rod to be calibrated is dropped into the core, and the resulting decay of neutron flux is observed and related to the reactivity. In this paper, the regulating rod will be calibrated according to the reactivity in OPR-1000 that corresponds to a certain control rod insert or withdraw, and the reactivity in power reactor depends on the integral and differential control rod group too. The core simulator OPR1000 is used to test those methods.

Analysis of Residence Time in the Measurement of Radon Activity by Passive Diffusion in an Open Volume: A Micro-Statistical Approach

Abstract: Residence time in a flow measurement of radioactivity is the time spent by a pre-determined quantity of radioactive sample in the flow cell. In a recent report of the measurement of indoor radon by passive diffusion in an open volume (i.e. no flow cell or control volume), the concept of residence time was generalized to apply to measurement conditions with random, rather than directed, flow. The generalization, leading to a quantity Δtr, involved use of a) a phenomenological alpha-particle range function to calculate the effective detection volume, and b) a phenomenological description of diffusion by Fick’s law to determine the effective flow velocity. This paper examines the residence time in passive diffusion from the micro-statistical perspective of single-particle continuous Brownian motion. The statistical quantity “mean residence time” Tr is derived from the Green’s function for unbiased single-particle diffusion and is shown to be consistent with Δtr. The finite statistical lifetime of the randomly moving radioactive atom plays an essential part. For stable particles, Tr is of infinite duration, whereas for an unstable particle (such as 222Rn), with diffusivity D and decay rate λ, Tr is approximately the effective size of the detection region divided by the characteristic diffusion velocity . Comparison of the mean residence time with the time of first passage (or exit time) in the theory of stochastic processes shows the conditions under which the two measures of time are equivalent and helps elucidate the connection between the phenomenological and statistical descriptions of radon diffusion.

Development of a LDIE Prediction Theory in the Condition of Magnetite Formation on Secondary Side Piping in Nuclear Power Plants

Abstract: It has been thought that wall thinning on the secondary side piping in nu-clear power plants is mostly caused by Flow-Accelerated Corrosion (FAC). Recently, it has been seen that wall thinning on the secondary side piping carrying two-phase flow is caused by not only FAC but also Liquid Droplet Impingement Erosion (LDIE). Moreover, it turns out that LDIE in nuclear power plants does not result from a single degradation mechanism but also from the simultaneous happenings of LDIE and FAC. This paper presents a comparison of the mass loss rate of the tested materials between carbon steel (A106 B) and low alloy steel (A335 P22) resulting from degradation effect. An experimental facility was set up to develop a prediction model for clarifying multiple degradation mechanisms that occur together. The experimental facility allows examining liquid droplet impingement erosion in the same conditions as the secondary side piping in nuclear power plants by generating the magnetite on the surface of the test materials. The magnetite is formed by controlling the water chemistry and the temperature of fluid inside the facility. In the initial stage of the experiments, the mass loss rate of A106 B was greater than that of A335 P22. However, after a certain period of time, the mass loss rate of A335 P22 became greater than that of A106 B. It is presumed that the results are caused by the different yield strengths of the test materials and the different degrees of buffer action of the magnetite deposited on their surfaces. The layer of magnetite on the surface of A106 is thicker than that of A335 P22, due to the different amount of chrome content. In nuclear power plants, carbon steel piping having experienced wall thinning degradation is generally replaced with low-alloy steel piping. However, the materials of pipes carrying two-phase flow should be selected considering their susceptibility to LDIE.

Gamma Radiation Measurements of Naturally Occurring Radioactive in Igneous Rocks and Its Radiological Complications

Abstract: The concentrations of natural radioactivity were measured in igneous rock samples collected from Albaha region in the south west of Saudi Arabia. A high purity germanium (HPGe) detector was used for analysis. The average activity concentrations for 226Ra, 232Th and 40K were 35, 31.52 and 843.63 Bq kg−1, respectively. The average absorbed dose rate was 70.86 nGy⋅h−1 with a corresponding average annual effective dose 0.09 mSv⋅y−1. The average radium equivalent activity value was 145.84 Bq⋅kg−1, lower than the international limit 370 Bq⋅kg−1. The external and internal indices average values were 0.39 and 0.49, respectively. The average results obtained in this study are lower than the average national and world recommended values, therefore, there is no health risk to the populace of the area. This study provides a baseline map of background radioactivity levels in the Saudi environment and will be used as reference information to assess any changes in the level background due to geological processes.

Generation and Detection of a Directed Monoenergetic Neutrino Beam with Hydrogen-Like Ions

Abstract: We consider two possible schemes for generation and detection of a monoenergetic directed beam of neutrinos which may have application to neutrino communication. First, we consider generation of a directed neutrino beam using electron capture beta decay in hydrogen-like ions. Next, we suggest detection of a directed neutrino beam using resonant absorption of a neutrino by a bare nucleus with the generation of a bound electron. This reaction is inverse to electron capture beta decay, and we call it “Bound State Inverse Beta Decay (BSIBD)”. We show that the recoil effect can be eliminated by an appropriate choice of velocities for the ions and bare nuclei. Finally, we consider a combination of a solid state source of a directed mono-energetic neutrino beam and its detection using BSIBD.

Determination of Uranium Traces in Nuclear Reactor IEA-R1 Pool Water

Abstract: IEA-R1 nuclear reactor operation has the routine to control uranium content in pool water to be in trace range below 50 μg/L. There are several routes to determine the uranium trace content in water in the literature; voltammetry has been systematically employed. In the present study, the chosen chemical determination of uranium traces used the voltammetric method known as AdCSV (adsorptive cathodic stripping voltammetry). This technique, based on mercury voltammetry, is an adequate methodology to determine uranium traces. The chloranilic acid [CAA] (2,5-dichloro-3,6-dihydroxy-1,4-benzo-quinone) is indicated as chelating agent. The redox reaction of UO2+2 with CAA is sensitive in the range of 2 2(CAA)2] reduction potential. In this work, we present the uranium trace results for IEA-R1 reactor water, sampled after an operation routine shutdown. The uranium trace determination for IEA-R1 pool water showed content around 1 μg/L [U] with statistical significance. Therefore the IEA-R1-reactor-water purification showed to be adequate and safe.

Activation Energy of Modified Peak Shape Equations

Abstract: The aim of this paper is to give some simplified expressions related to the peak shape method. The modified equations have been used to calculate the activation energy (E) of commercial thermoluminescent dosimeters (TLD), as well as of ZnO thermoluminescent material produced in laboratory; the values so determined have been compared to the values obtained using the classical expressions of the peak shape method. The modified equations proposed are as a function of peak shape parameters or the peak temperature at the maximum. This expression could be useful to obtain approximated E values in the case of complex glow curves as well, when the peaks are not well resolved but the peak temperature at the maximum may be easily determined.

Numerical Analysis for Transients in External Source Driven Reactors

Abstract: The main purpose of this paper is to perform a numerical analysis of the Neutron Spatial Kinetic Equations, subject to transients of the External Neutron Source, by applying the Implicit Euler Method as well as the Runge-Kutta Method in order to check which methods are best applicable in transients caused by External Neutron Source. For this purpose, a one-dimensional ADS reactor with a constant external source was simulated based on the geometry of ANL-BSS-6 reactor for benchmark effects.

Transient Studies of Ghana Research Reactor-1 after Nineteen (19) Years of Operation Using PARET/ANL Code

Abstract: The Program for the Analysis of Reactor Transients/Argonne National Laboratory (PARET/ANL) code was used to predict the thermal hydraulic behaviour of the Ghana Research Reactor-1 after adding 9.0 mm of beryllium to the top shim tray of the core. The core was analysed for reactivity insertions 2.1 mk, 3.0 mk, 4.0 mk, 5.0 mk and 6.7 mk, respectively. The reactor is still safe to operate in the range 2.1 mk to 4.0 mk. However, 2.1 mk would be ideal since the reactor automatic shutdown (SCRAM) is set not to exceed 120% of reactor nominal power.

Studying the Nuclear Structure of Some Target Nuclei Used for Radiotherapy Nuclei Production by Using Skyrme-Hartree-Fock Method

Abstract: The nuclear structure for some target nuclei namely: 32S, 58Ni, 89Y, 90Zr, 100Mo and 197Au used for production of the therapeutic radionuclides; 32P, 58Co, 89Sr, 90Y, 99Mo, 100Tc, 197Pt and 197Hg has been investigated using Skyrme-Hartree-Fock method based on Skyrme effective two-body interaction. For these purpose, we have calculated the various nuclear densities, the corresponding root mean square radii and nuclear binding energies. The density dependent initial neutron and proton exciton numbers have been also calculated which give the ability to investigate the neutron and proton induced reaction cross-sections for these target nuclei using hybrid model for pre-equilibrium nuclear reactions. The calculated results are compared with available experimental data.

Geometric Models of Atomic Nuclei

Abstract: While experimenting with the more and more popular neodymium magnetic ball sets, the author developed a method, by which models of atomic nuclei can be created. These macroscopic models visually represent several features of nuclei and nuclear phenomena, which can be a useful mean during the teaching of nuclear physics. Even though such macroscopic models are unable to depict the true quantum physical nature of nuclear processes, they can be much more useful didactically than the previously used disordered sets of balls, to represent the atomic nucleus.

Electrostatic Theory of Elementary Particles

Abstract: Theoretical physics makes a wide use of differential equations for which only a potential solution is applied. The possibility that these equations may have a non-potential solution is ruled out and not considered. In this paper an exact non-potential solution of the continuity equation is described. The electric field of an elementary charged particle consists of two components: the known Potential Component (PC) produced by the charge and the earlier unknown Non-potential Component (NC) with a zero charge. Charged particles have both components, while a neutron has only the NC. The proton and neutron NC ensures similarity of their properties. The PC is spherically symmetric and NC is axisymmetric. Therefore, to describe an elementary particle, one should take into account both its spatial coordinates and the NC orientation. The particle interaction is determined by their NC mutual orientation. Neglecting the latter leads to indefiniteness of the interaction result. In a homogeneous electric field, the force acting on the NC is zero. Therefore, a charged particle possessing the NC will behave like a potential one. In an inhomogeneous field, the situation is principally different. Due to the NC there occurs an interaction between a neutron and a proton. The non-potential field results in the existence of two types of neutrons: a neutron and an antineutron. A neutron repels from a proton ensuring scattering of neutrons on protons. An antineutron is attracted to a proton leading to its annihilation. The NC produces the magnetic dipole moment of an elementary particle.

Coupling a Gas Chromatography Unit to an Inductively Coupled Plasma Mass Spectrometer

Abstract: Although the eminent threat of a terrorist group detonating an improvised nuclear device (IND) in an urban environment is low, it is crucial that countries develop modern nuclear forensic capabilities to expedite response in a post-detonation scenario. In particular, new instruments need to be created to shorten dissolution time, expedite chemical separation, and improve forensic analysis of the nuclear melt glass that is created during the detonation of the device. To expedite this process, an instrument was designed to thermally couple a gas chromatograph (GC) to a time-of-flight inductively coupled plasma time-of-flight mass spectrometer (ICPTOFMS) In order to couple these two instruments, another instrument was designed to provide an isothermal atmosphere between the GC and TOFICPMS to expedite rapid gas separations processes. By using gas separations instead of the current wet chemistry processes, the required separation and analysis time of the melt glass significantly decreases. The new instrument would also provide a more detailed analysis of the elemental and isotopic composition of the melt glass. By completing these tasks simultaneously, this significantly decreases the required time to conduct these separations and improves the elemental and isotopic analysis.

An Integrated Process for Recycling of ThO 2 Based Mixed Oxide Rejected Nuclear Fuel Pellets

Abstract: This paper presents a study on the process engineering aspects of relevance to the industrial implementation of ThO2 and (Th, U)O2 mixed oxide (MOX) pellet type fuel manufacturing. The paper in particular focuses on the recycling of thoria based fuel production scrap which is an economically important component in the fuel manufacturing process. The thoria based fuels are envisaged for Advanced Heavy Water Reactor (AHWR) and other reactors important to the Indian Nuclear Power Programme. A process was developed for recycling the chemically clean, off-specification and defective sintered ThO2 and (Th, U)O2 MOX nuclear fuel pellets. ThO2 doesn’t undergo oxidation or reduction and thus, more traditional methods of recycling are impractical. The integrated process was developed by combining three basic approaches of recycling namely mechanical micronisation, air oxidation (for MOX) and microwave dissolution-denitration. A thorough investigation of the influence of several variables as heating method, UO2 content, fluoride and polyvinyl alcohol (PVA) addition during microwave dissolution-denitration was recorded on the product characteristics. The suitability evaluation of the recycled powder for re-fabrication of the fuel was carried out by analyzing the particle size, BET specific surface area, phase using XRD, bulk density and impurities. The physical and chemical properties of recycled powder obtained from the sintered (Th1-y, Uy)O2 (y; 0 - 30 wt%) pellets advocate 100% utilisation for fuel re-fabrication. Recycled ThO2 by integrated process showed distinctly high sinterability compared to standard powder evaluated in terms of surface area and particle size.

Sub-Criticality Measurement with Source Term for Research Reactor in Inverse Kinetics Method

Abstract: In reactor physics tests, it is important to monitor sub-criticality continuously during criticality approach. Reactivity measurements by the inverse kinetics method are widely used during the operation of a nuclear reactor. This technique is successfully applied at sufficiently high power level or to a core without an external neutron source where the neutron source term in point reactor kinetics equations may be neglected. For operation at low power levels or in the sub-critical domain, the increase in the fluctuation of the neutron signal may cause difficulties in the evaluation of reactivity and the effect of direct emission from the external neutron source may not be neglected. Therefore, contribution of the neutron source must be taken into account and this implies knowledge of a quantity proportional to the source strength, which calls the source term and then it should be determined. The research work has been conducted to measure reactivity with source term using a dedicated reactivity measurement system by the Least Square Inverse Kinetics Method (LSIKM). Application to a simulator of HANARO research reactor, Korea Atomic Energy Research Institute (KAERI), with known source strength and reactivity worth has showed consistent and satisfactory agreement.

Further Studies of BURSTS and Spallation in High-Energy Heavy Ion Reactions

Abstract: Aspects of BURSTS and Spallation reactions induced by high-energy heavy ions in thick targets (>10 cm thick) will be investigated: BURSTS are reviewed from a historical and phenomenological point-of-view. Details of interactions in nuclear emulsions will be compared for irradiations of 72 GeV 22Ne-ions from Dubna with irradiations of 72 GeV 40Ar-ions from Berkeley. Measured correlations in individual interactions between multiplicities of “minimum ionizing particles”, ns, and “black prongs”, nb, will be shown as “ns-vs.-nb” per event for BURSTS and separately for Spallation in interactions of 72 GeV 22Ne-ions. Monte Carlo calculations, based on the MCNPX 2.7 code, have been carried out for 72 GeV 22Ne interacting in nuclear emulsions: The correlation between ns and nb in Spallation reactions could be understood. However, “ns-vs.-nb” correlations from BURST-interactions could not be reproduced with this model for events with small numbers of heavy prongs nh ≤ 10. For large numbers of heavy prongs with nh > 10 one could find some agreement between experiments and calculations, however, not in all details. Further experimental and theoretical studies are necessary before one has a complete understanding of BURST interactions in high-energy heavy ion reactions.

On-Site Calibration Method of Dosimeter Based on X-Ray Source

Abstract: The real-time monitoring of environmental radiation dose for nuclear fa-cilities is an important part of safety, in order to guarantee the accuracy of the monitoring results regular calibration is necessary. Around nuclear facilities there are so many environmental dosimeters installed dispers-edly, because of its huge quantity, widely distributed, and in real-time monitoring state; it will cost lots of manpower and finance if it were tak-en to calibrate on standard laboratory; what’s more it will make the en-vironment out of control. To solve the problem of the measurement ac-curacy of the stationary gamma radiation dosimeter, an on-site calibra-tion method is proposed. The radioactive source is X-ray spectrum, and the dose reference instrument which has been calibrated by the national standard laboratory is a high pressure ionization. On-site calibration is divided into two parts; firstly the energy response experiment of dosim-eter for high and low energy is done in the laboratory, and the energy response curve is obtained combining with Monte Carlo simulation; sec-ondly experiment is carried out in the field of the measuring dosimeter, and the substitution method to calibrate the dosimeter is used; finally the calibration coefficient is gotten through energy curve correction. In order to verify the accuracy of on-site calibration method, the calibrated dosimeter is test in the standard laboratory and the error is 3.4%. The re-sult shows that the on-site calibration method using X-ray is feasible, and it can improves the accuracy of the measurement results of the stationary γ-ray instrument; what’s more important is that it has great reference value for the radiation safety management and radiation environment evaluation.

Effect of Kinetic Parameters on I-129 Activity from Fuel to Coolant in PWRs

Abstract: Effect of kinetic model parameters on fission product (I-129) activity from fuel to coolant in PWRs has been studied in this work. First a computational model was developed for fission product release into primary coolant using ORIGEN-2 as subroutine. The model is based on set of differential equations of kinetic model which includes fuel-to-gap release model, gap-to-coolant leakage model, and Booths diffusion model. A Matlab based computer program FPAPC (Fission Product Activity in Primary Coolant) was developed. Variations of I-129 activity in Primary Heat Transport System were computed and computed values of i-129 were found in good agreement and deviations were within 2% - 3% of already published data values. Finally, the effects of coolant purification rate, diffusion constant and gas escape rate on I-129 activity were studied and results indicated that the coolant purification rate is the most sensitive parameter for fission product activity in primary circuit. For changes of 5% in steps from −10% to +10% in the coolant purification rate constant (Β), the activity variation after 200 days of reactor operation was 23.1% for the change.

Neutron Flux Signal Acquisition from Plant Instrumentation Channel of Research Reactor for Reactivity Calculation

Abstract: A design for instantaneous neutron flux signal acquisition system is being carried out for reactivity measurement of the nuclear research reactor. It is a computer-based digital data acquisition system that can perform continuous monitor and measurement of reactivity inserted into or removed from the research reactor. The acquisition system accomplishes with two major parts. The first part is an interfacing PCI based data acquisition card and the corresponding driver software intending to on-line acquisition of neutron flux signals from plant instrumentation channel. The second part incorporates the high-level Visual Basic real time program, indigenously developed for computation of reactivity by the solution of neutron point kinetic equations and other relevant functional modules like input file logging, reactivity calculation, graphics demonstration etc.