Nader M. A. Mohamed

Partial Loading of Thorium-Plutonium Fuel in a Pressurized Water Reactor

Abstract This research focuses on the utilization of thorium-plutonium fuel in pressurized water reactors (PWRs). The reference PWR selected in this research was the Westinghouse AP1000. Thorium-plutonium mixed-oxide (MOX) fuel assemblies partially replaced the uranium oxide fuel assemblies to reduce uranium demand. The cases studied contained 36, 48, 60, 72, and 84 thorium-plutonium MOX fuel assemblies, with the rest of the 193 fuel assemblies loaded with UO2 fuel. The core cycle length, the amount of plutonium incinerated, the amount of generated 233U in the spent fuel, and the conversion ratios were determined using MCNP6. For the different cases, safety parameters such as the power peaking factor and delayed neutron fraction (βeff) were evaluated. The study showed that using thorium-plutonium MOX can achieve good peaking power factors with delayed neutron fractions within the safety limits. Also a conversion factor of about 10% was achieved.

Image enhancement using MCNP5 code and MATLAB in neutron radiography.

This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work.

Study of Using Zirconium as a Reflector for Light Water Reactors

Abstract Zirconium is studied in this paper as a reflector for light water reactors. An exploratory analysis of using zirconium as a reflector for two simple reactor core models was carried out. The study showed that use of zirconium as a reflector has a valuable impact on the core reactivity. The study also showed that zirconium-water reflector is more effective than water reflector or stainless steel-water reflector. A typical Westinghouse 1150-MW(electric) pressurized water reactor was simulated using the Monte Carlo code MCNP5 as a case study. The simulation was carried out at the beginning of the core cycle of three batch cores with 235U enrichments of 2.25, 2.8, and 3.3 wt%. The simulation showed that use of Zircaloy-4 reflector between the fuel assemblies and the core barrel adds a positive reactivity Δkeff of 0.00686, while use of stainless steel reflector adds a positive reactivity Δkeff of 0.0037. Use of Zircaloy-4 reflector increases the relative power density in the peripheral assemblies by ˜38%. The power peaking factor is shifted from the center toward the periphery, and the assembly power peaking factor is reduced by ˜13%. The use of Zircaloy-4 reflector with this increase of the reactivity of the peripheral assemblies increases the fast neutron current (E > 0.5 MeV) that reaches the reactor pressure vessel (RPV) by 70%, while the use of stainless steel reflector reduces it by 44%. Adjusting the 235U enrichment in the peripheral assemblies batch to compensate for the excess reactivity caused by using Zircaloy-4 reflector reduces the 235U enrichment by 8.5% in this batch. This means a reduction of 3.35% of the core 235U average enrichment can be achieved by the use of Zircaloy-4 reflector. This reduction in the 235U enrichment reduces the increase of the fast neutron current that reaches the RPV to 23%. In this case, increasing the water gap between the core barrel and the RPV by 3 cm reduces the fast neutron current that reaches the RPV to 95% of that of the basic case. The use of Zircaloy-4 reflector has a good effect on flattening the fission density distribution in the peripheral assemblies batch both before and after reducing 235U enrichment.

Evaluation of the Efficiency of Gamma Spectrometers for Measuring Volumetric Samples

Abstract A new concept was proposed to calibrate gamma spectrometers for measuring the activity of volumetric samples. The concept used a formula that was developed to give a correction factor to the detector efficiency when the detector was calibrated with a point source. The concept gave efficiency curves for an HPGe detector. The results obtained by the concept yielded deviations between 0.5 to 5% from the experimental data. The concept needs only the availability of the MCNP code or similar gamma transport codes.

Detection of illicit material using neutron activation: weakness and solutions

Poisoning the illicit materials by a neutron absorber leads to false detection when the detection is relied on combined thermal neutron activation and fast neutron activation to identify the elements of interest. The use of adjacent transmission thermal neutron detector for verifying the presence of neutron poisons and to trigger an alarm was investigated experimentally and using MCNP calculations. The illicit material of high hydrogen content will affect the detector response in the presence or absence of poisons.

Towards a methodology for bulk sample neutron activation analysis

Abstract The main challenge in large sample neutron activation analysis (LSNAA) is the determination of neutron self-shielding and gamma ray self-attenuation corrections. After these corrections are determined, the analysis proceeds as in normal neutron activation analysis (NAA), as if the sample were infinitely small. In this paper, these corrections are calculated using the MCNP code for different standard sample geometries with different diameters. Modelling studies for LSNAA using an external neutron beam were performed. An analytical formula for the correction factors for neutron self-shielding and gamma ray self-attenuation is derived. The correction factors as well as flux parameters are calculated analytically. The analytical formula is verified using the MCNP code. All of the calculated parameters were tabulated and graphed. From the calculated data, other unknown material parameters could be obtained based on tabulated data or graphs. This method is a direct and easy method to perform large sample neutron activation analysis without complex calculations. In addition, for the user who does not have good experience with codes such as MCNP, she/he can use the chart or the tabulated information to define their unknown sample with the required information for the LSNAA experiment.

Correction to: Concept design of an illicit material detection system

In the original publication of the article, the Acknowledgement section was missed. The complete Acknowledgement is given in this correction.

Monte Carlo simulation of the responses of gaseous effluent monitors to radioactive isotopes

This paper discusses use of Monte Carlo simulations to facilitate testing and calibration of a gaseous effluent monitor (GEM). MCNP5 was used to simulate responses of the 131I and 41Ar GEM detecting units exposed to specific radioactive sources. The agreement between the MCNP5 predictions and experimental measurements is good enough to validate the MCNP5 model. It has been demonstrated that the Monte Carlo code is a powerful and useful tool to determine accurate detector responses and facilitate the calibration process of this type of the monitors.

Application of k 0 -INAA for the determination of essential and toxic elements in medicinal plants from West Pokot County, Kenya

The k0-INAA has been applied for determining nineteen elements in six medicinal plants used to cure various diseases in West Pokot, Kenya. It was observed that some elements are accumulated in excessive amount by the tested plants. Care has to be taken to control the daily dose of these herbs. Reference materials (IAEA-155 and IAEA-V-10) were analyzed simultaneously with the samples to evaluate the accuracy of the analysis protocol. It was found that the measured elemental concentrations were in a good agreement with the reported values with a maximum deviation of 20%.

DECREASING THE DETECTION LIMIT OF COPPER FOR THE INSTRUMENTAL NEUTRON ACTIVATION ANALYSIS BY MEASURING THE ACTIVITY OF 64Cu ISOTOPE FROM THE ANNIHILATION PEAK

Abstract A procedure was developed for measuring the concentration of copper, in the Instrumental Neutron Activation Analysis method, by measuring the produced 64Cu isotope activity (after irradiation) from the annihilation peak (511-keV peak). In this procedure the number of counts under the annihilation peak is divided into two categories: (a) counts coming from the decay of the 64Cu isotope and (b) counts coming from the interactions of energetic photons (with energies >1.022 MeV, the pair production threshold) with the detector and surrounding materials. The last category is evaluated and subtracted from the annihilation peak counts, and the rest of the counts are used to calculate the activity of 64Cu. Measuring copper concentration using this method will improve its detection limit. The method was validated by measuring the concentration of copper in four International Atomic Energy Agency (IAEA) reference materials: Soil-7, Lake Sediment, Human Hair, and Hay Powder. The maximum deviation between the results and that given in IAEA certificates is 4.4%. The method decreased the detection limits of the four samples to ˜3, ˜4.5, ˜0.6, and ˜1 mg/kg, respectively.