Study of the natural radioactivity background of the Douala University Campus I and II by gamma spectrometry

L’objectif du travail de Master II était d’évaluer superficiellement le niveau radiologique des sols des campus I et II de l’Université de Douala par spectrométrie gamma. Comme il existe différents types de spectromètres gamma nous avions utilisé un détecteur Germanium hyper pur et haute sensibilité, à large gamme d’énergie, « Broad Energy Germanium Detector » (BEGE 6350 model) du Laboratoire de spectrométrie de l’ANRP. La particularité de cet appareil est qu’il a la capacité de détecter les faibles gammes d’énergie propre aux radionucléides naturels tels 226Ra, 232Th et 235U. L’analyse des spectres des échantillons de sol a été faite avec le logiciel d’analyse Genie-2000 version 3.2 (dernière version) qui prend en module le paquet de calibration LabSOCS. La calibration a été faite avec des sources de références fournies par l’AIEA et des sources étalons préparées au sein même du laboratoire et compléter par le module de simulation LabSOC. Afin d’évaluer l’exposition aux rayonnements ionisants, certains paramètres radiologiques ont été estimés. Ils ont été comparés avec les limites de sécurité recommandées par l’AIEA et « UNSCEAR ». Un total de dix-huit (18) échantillons de sol a été prélevé (07 du sol du campus 1 et 11 du sol du campus 2). Ce prélèvement tenait compte d’un échantillonnage mixte sur le terrain afin d’avoir une bonne représentativité par les résultats finaux. Le nombre d’échantillons prélevés par campus tenait compte de la superficie des différents campus. La préparation et l’analyses des 18 échantillons s’est étendue sur une période de 06 mois avec vérification des premiers résultats par une seconde analyse suivies par le dépouillement des spectres et l’interprétation des résultats. Le profil des activités spécifiques des radionucléides présente une faible activité dans les zones étudiées seulement dans le cas du potassium. Les valeurs moyennes obtenues pour le 226Ra, le 232Th et le 40K dans les deux campus sont respectivement : 25.48Bq/kg, 65.96Bq/kg et 39.14Bq/kg pour les échantillons du Campus 1 et 24.50Bq/kg, 66.71Bq/kg et 28.19Bq/kg pour les échantillons du Campus 2. Les valeurs moyennes de l’activité équivalente de radium sont 122.81Bq/kg et 122.08Bq/kg. Les valeurs moyennes estimées et calculées sont de 99.13nGy/h et 98.18nGy/h pour le débit de dose absorbée dans l’air, 0.12mSv/an et 0.12mSv/an pour la dose efficace annuelle en plein air, 0.34 et 0.33 pour le paramètre de risque externe dans le Campus 1 et le Campus 2 respectivement. Les valeurs des paramètres évalués sont en dessous des valeurs limites de sécurité recommandées par « UNSCEAR » 2000, sauf le débit de dose absorbée dans l’air et la dose efficace annuelle en plein air qui sont plus élevés par rapport aux valeurs de 60nGy/h et 0,07mSv/an. Tout en restant vigilant sur ces valeurs de doses élevées, le calcul du risque radiologique externe nous donne de croire que la communauté universitaire est à l’abri du danger des rayonnements ionisants. Cependant, des investigations futures avec un échantillonnage plus réduit feront l’objet d’études et de validation de ces résultats qui ont été transmis à l’agence nationale de radioprotection

SHIELDING DESIGN AND SAFETY MEASURES AROUND 60Co, 192Ir, AND 252Cf SOURCES IN INDUSTRIAL RADIOGRAPHY FACILITIES BASED ON PHITS MONTE CARLO SIMULATIONS

Industrial radiography has been spreading worldwide and its use as a non-destructive testing method for different equipment, goods, and objects of all sorts has proven to be an essential and indispensable technique for objects’ control. Depending on the type of sources used, attention must be paid differently on the protective measures to be taken around a facility built for radiography purposes. In this regard, protection measures around an X-ray generator are different from those necessary to protect people and the environment around a radioisotope, which spontaneously emits radiation, alpha, beta, gamma, or neutron, until it decays to a stable element. As great issues of protection highlighted along with justified activities are the measures to limit the dose to an acceptable limit, the present study was projected to solve the problem of shielding design around industrial radiography facilities using gamma and neutron sources. Gamma sources of Co-60 under 100 Ci and Ir-192 under 50 Ci and a neutron source of Cf-252 with an activity of 100 Ci were considered as typical sources used in fixed industrial radiography. The PHITS Monte Carlo code was used to compute the appropriate concrete wall necessary to shield gamma radiation and the most effective material appropriate to shield neutron source. For the gamma radiation source, the shielding design took into consideration the double and single corner maze, while the neutron radiography testing room design considered only a reinforced shielded door, without maze design. The result of the calculation showed that the minimum concrete wall thicknesses necessary to shield the radiation were 70 cm, 120 cm, and 250 cm for the Ir-192, Co-60, and Cf-252 sources, respectively. All the presented values were obtained with acceptable statistic except the case of Cf-252 where the calculation involving neutron interaction was time-consuming and computing resources demanding. In addition, different materials were subjected to investigation to find out the most appropriate for neutron shielding as the Cf-252 spontaneous fission produce fast neutrons. These fast neutrons undergo moderation in hydrogenous or light material prior to their absorption in borate material where they produce gamma-rays, that are shielded using concrete of dense material (high Z material). It was found that tungsten was efficient in shielding energetic neutron, but is expensive to be used in shielding design. Paraffin, graphite, polyethylene, plastic, and iron could be used for the neutron slowing -down process and the combination of some of these materials is more effective. Water (light and heavy), the most effective neutron moderator or lithium cannot be used in radiographic test rooms as they posed a structure problem (easy leakage, difficult to use in wall form). Also, while using hydrogenous materials, attention should be paid to the safety and security of the design, construction, and operation of the facility as the fire likelihood increases. Gamma-rays shielding design showed that the most appropriate design to shield gamma from sources used in industrial radiography facilities is the double corner maze compared to the single corner maze. The maze length necessary to shield gamma radiation up to 2.5 µSv/h in single corner maze design is almost the triple of the maze length needed in double corner maze design. The length is cost-related. This conclusion highlighted the importance of cost-benefit analysis in the early planning stage prior to its design, construction, commissioning, operation, and decommissioning. Three different positions of the source were evaluated and the outcome showed that the central position of the source in radiography testing facilities should never be allowed unless there are special justifications: this position was found to be the case with the highest exposure rate to the boundary areas enclosing the facility. The choice between the left and right position of the source in the facility merely depends on the length of the maze in the case that the design is a single corner maze. For example, in the case that the source of cobalt-60 is used for neutron imaging, the left position of the source should be preferred when the maze length is less than 7 m and the right position if not. For iridium-192 source in industrial radiography facilities, the single maze design showed that at least 5 m maze length is needed to set the source position on the right side. Neutron shielding is complex in industrial radiography using fast neutrons as those emitted by the spontaneous fission source of Cf-252 investigated in this research. Maze design is not appropriate for neutron shielding and the shielded enclosure, in this case, should be made of reinforced door including materials for neutron slowing down, its absorption, and the gamma-ray shielding. From the computation based Monte Carlo methods, the following materials should be used in combination with iron for slowing the energetic neutron, boron or borate materials for neutron absorption, and concrete or lead at the end of the chain for gamma attenuation. The combination of three to five materials could be enough when well-chosen and the shielded enclosures should be designed and constructed based on a long term planning evaluation. Attention should be paid on materials used to avoid the critical problem of activation as the neutron interaction with an alloy is likely to produce new radioisotope with a high risk of contamination. To keep the effective dose rate as low as reasonably achievable (ALARA) in the areas adjacent to the radiographic facilities, the International Atomic Energy Agency (IAEA) has developed regulatory guidelines, safety reports series, and a various number of technical documents to help governments, institutions, and individuals involved from near or far in the use of radioactive sources in the practice of radiography. These guidelines have been used worldwide to develop a safety culture and network with less risk of radiological accidents from the industrial side. As the National Radiation Protection Agency of Cameroon is the regulatory authority in Cameroon for the practice of industrial radiography, there is a particular need for clear and precise regulations on the use of radioactive sources for non-destructive tests. The regulatory authority has the opportunity to consider the Korean regulations (or USNRC one) and the IAEA case to develop appropriate laws and rules for safe acquisition, possession, planning, siting, designing, construction, commissioning, operation, and decommissioning of a radiography testing facility. The licensees, radiographers, and any individual involved in the use of gamma or neutron sources for activities related to industrial radiography could then have a law as foundation and regulation for their practices without ambiguity

RADIATION PROTECTION OPTIMIZATION IN FIXED INDUSTRIAL RADIOGRAPHY-BASED PHITS MONTE CARLO CODE SIMULATION

Nondestructive test (NDT) has been widely used for defects detection, well joints inspection, and material integrity verification in recent decades. Gamma radiography’s applications in industry are known as high risk-related techniques among the nuclear-related industries nowadays. For example, the IAEA safety report about the lesson learned from accidents in industrial radiography reported that around 45 % of accidents in the nuclear industry accounted for industrial radiography in both developed and developing countries [1]. This demonstrates the need to optimize safety and protection measures around gamma sources used in fixed industrial radiography. The present study focuses on the optimization of engineering barriers during the sitting and construction phases of such facilities. Monte Carlo methods-based PHITS are used to optimize the shielding design of enclosures of a radiographic facility to create a safe working environment for both radiographers and the public. The Particle and Heavy Ion Transport code System [2, 3] was used to determine the appropriate concrete wall and dose estimation to keep radiation As Low As Reasonably Achievable (ALARA). As shown in Fig.1, there are many concerns in old built facilities and few in new ones, but the primary objective is to provide radiological protection regulation or to update and optimize the existing regulation, taking into consideration the shielding design of the facility As the radiation exposure to any given material depends on the thickness of the shielding, the quantity, and the energy of the radiation, both Co-60 and Ir-192 sources were used in our study as they are the most used high-energetic radionuclides used in industrial radiography. The MC code used for computation, PHITS, is a general-purpose Monte Carlo Particle and Heavy Ion Transport code System developed by a collaboration between Japanese institutions and Europe. The version used, PHITS 3.10 with several changes, allows the simulation of photon and other particles of interest transport over a wide range of energy. The design principle here is based on providing enough shielded enclosure to keep the dose rates out of the facility (in the closest adjacent areas to the facility) lower than 2.5 μSv/h, in adherence to the ALARA principle for the public exposure. If not, a large exclusive area should be set, but this part is considered as administrative controls, which are discussed differently. The facility design was based on the IAEA safety report series [7] and the photon flux distribution simulated is presented in Fig. 2 along with the plot of the dose conversion factor used for computation. Appropriate concrete thickness to shield radiation from isotropic sources was computed prior to the radiographic testing room design. The minimum concrete wall thickness for the facility using Co-60 sources described previously was found to be 120 cm while it was found to be 70 cm for Ir-192 related facility as shown in TABLE 2. To ensure that the radiation dose falls under the recommended limits set either by international organizations as IAEA, or by the regulatory authority of Cameroon, the necessary and appropriate shielding walls and shielding materials shall be installed and regulated by laws. It is recommended to the government, conjointly with the NRPA of Cameroon, to make a law project in this regard that will be passed in view to facilitate the regulation of the wide-spreading practice of industrial radiography using radioactive sources, especially gamma imaging. The present code could be really helpful to the Government of Cameroon in developing a database for sitting and construction of radiographic testing rooms depending on the radionuclide type, its energy and intensity, and its activity. Appropriate design-based maze technology was developed by Guembou in his thesis [6]. There is a real need for the implementation of the international rules and the adoption of clear and specific national guides for the application of industrial radiography in Cameroon as well as in different other countries. Different IAEA safety standards series, safety reports series, and technical documents were provided to help governments, institutions, and individuals involved in the use of radioactive sources for industrial radiography to develop an appropriate safety culture. In this regard, enclosures of a radiographic room should be properly designed and used for the sources for which they were designed, considering the maximum activity, the type of radioisotope, their energy, and intensity. Developing countries as Cameroon could use MC simulation for national regulation improvement according to their socioeconomic statute and technology-based considerations

Study of the natural radioactivity background of the Douala University Campuses and surrounding by Nuclear Techniques: Validation by GEANT4 Monte Carlo simulations

The present thesis focused on an ionizing radiation monitoring project of the soils of the University of Douala campuses (campus 1 and 2) and surrounding, the Littoral region of Cameroon. The purpose was to provide a baseline to document the conditions present at the time of sampling. The methodology used γ-ray spectrometry-based High Purity Germanium detector (HPGe), both Energy Dispersive X-Ray Fluorescence (EDXRF) and Wavelength Dispersive X-Ray Fluorescence (WDXRF) spectrometry for sample’s elemental characterization, and the Monte Carlo simulation-based Geant4 toolkit for detector efficiency calibration. The Geant4 toolkit also provides the opportunity to optimize the detection systems using computer simulations and greatly reduces the need for expensive (radiation exposure to calibration sources) testing in the laboratory. The assessment of 238, 235U, 232Th, 137Cs, and 40K concentration was done by measuring soil and sand samples by γ spectrometry-based High Purity Germanium detectors (HPGe). Both laboratories of the National Radiation Protection Agency (NRPA) of Cameroon and the Atomic and Nuclear Spectroscopy, Archeometry Laboratory of the University of Liege were used for experiments. Geochemical characterization of soil samples, origin determination, and provenience were accessed by X-ray spectroscopy. By comparing the results of two detectors and the technics used according to the detector type, improvements on the γ spectrometry methodology were made. The relative uncertainty activity concentration was calculated for 226Ra, 232Th, and 40K. The average report between the GC0818-7600SL model and the BEGe-6530 model was the main outcome that suggested real attention that should be paid when selecting the radionuclide to be investigated on a specific type of detector. The BEGe detector was found to be more suitable for low γ energy emitters measurement, compared to the GC0818-7600SL model, found more efficient for high energy γ emitters. The potential radiological hazards parameters were assessed by calculating successively the following parameters from using those sands in the construction of dwellings and large buildings: Ra-dium Equivalent activity (Raeq), Outdoor absorbed γ dose rate (Dout), Annual Effective Dose rate (AED), Internal hazard (Hin) and external hazard (Hex) indexes, and and γ indexes for sand samples used as building materials. Results obtained show that Annual Effective Dose absorbed by in-habitants due to the use of the investigated sand as construction materials was found to be below 1.0 mSv y-1. Therefore, sand used as building materials from the investigated quarries appears to be radio-logically safe for building construction and for the environment (beaches, built houses, …) where people could safely spend time. Soil characterization using EDXRF in the present study provided an overview of the geological origin or provenience of the investigated area. As a result, the analyzed soil samples could be classified chemically as Fe-soil and are illustrative dregs from the Continental margin because of the high concentration of Fe in all the investigated samples. These data record the elemental composition and the natural radionuclide’s radioactivity levels of the studied area and could be set as reference database information in the region, in Cameroon as well as for the Gulf of Guinea’s data. Monte Carlo validation based on the GEANT4 toolkit has been used to validate the efficiency calibration of the system, and it has been noticed that the combination of γ-ray spectrometry, the development of related Monte Carlo methods, and the GEANT4 (GEometry ANd Tracking) toolkit developed for γ spectrometry simulation are compelling and useful for detector characterization nowadays. It can then be concluded that the Monte Carlo simulation gives more prominent adaptability, greater flexibility, gained time, precision, and accuracy when determining the detector response and efficiency in the routine of environmental radioactivity monitoring

Shielding design for high-intensity Co-60 and Ir-192 gamma sources used in industrial radiography based on PHITS Monte Carlo simulations

Protection in industrial radiography facilities has been a worldwide concern for decades and the most appropriate protection for the public and operators in small-scale facilities is the engineering protection including shielding first, time and distance. Monte Carlo methods were used for thickness determination in shielding design for gamma imaging using gamma sources in fixed facilities. Computation was done to develop the concrete design that can slow down the effective dose rate of the gamma rays from 100 Ci (or less) of Co-60 and 50 Ci (or less) of Ir-192 less than the value of 2.5 μSv/h (the limit for Public area) as required by the ALARA principle and regulation of different countries. The minimum concrete wall thickness necessary to achieve the ALARA principle in the previous conditions was found to be 120 cm for Co-60 source and 70 cm for the Ir-192 source. From the optimized design using a single corner maze, it was found that the central source position gives rise to high exposure and should not be used as a source position. Only the left and right positions of the source are preferred depending on the facility’s dimensions. For Co-60 with a maze corridor less than 7 m, the right position of the source is the most appropriate while the left position is preferred for larger dimensions. For Ir-192, the right position if preferred if the corridor length is less than 5 m and the left otherwise. Double-corner maze design was found to be the most appropriate shielding design for gamma radiation with the left and right position of the sources the best for both sources used. The double-corner maze required only 4 m maze length to achieve 2.5 µSv/h or less ALARA principle while single corner design requires 12 m or 14 m for left and right positions, respectively

Maze influence to radiological protection around industrial radiographic sources (Co-60) under 100 Ci

peer reviewedThe shielded enclosure design around the gamma radiography facilities under 100 Ci cobalt-60 source was evaluated as well as the maze design and source positions contribution to the dose limitation consistent with the ALARA principle. It was found that the most effective maze type to shield gamma radiations was double (multiple by extension) corners maze type. From discussions on the source positions, practitioners should select the optimizing position from both left or right depending on the length of the maze in the case of a single corner type, but never at the central position. The obtained results provided an insightful contribution to the radiological protection in industrial radiography

Counting time measurement and statistics in gamma spectrometry: the balance

Nuclear counting statistics at high count rate are assessed on a γ-ray spectrometer set-up. Our typical gamma spectrometry system consists of a High Purity Germanium (HPGe) detector, liquid nitrogen cooling system, preamplifier, detector bias supply, linear amplifier, analog-to-digital converter (ADC), multichannel storage of the spectrum, and data readout devices. Although the system is powerful enough for background measurements, it is important, nowadays, to have a great statistical in short time measurement: which is a challenge for scientists. The purpose of this study was to determine the average time for gamma spectrometry measurement. To detect Uranium, Thorium and their respective daughters and Potassium series with a relative related error less than 1%, it was found that it is necessary to count during a minimum of 24 Hours (86,400 s). This result is in accordance to the literature with planar geometry detector. These results conduct us to make the following three guidelines for selecting the detector best suited for an application: 1. The more detector material available (germanium semi-conductor), the higher the full-energy peak efficiency. 2. The smaller the distance between the detector and the source material, the higher the full- energy peak efficiency. 3. While better resolution gives a better MDA, the resolution contributes only as the square root to the MDA value, whereas the MDA is proportional to the full-energy peak efficiency. This idea came to us by comparing the spectra of measuring radioactivity lasts for 12 hours in the day that does not fully covered the night spectra for the same sample. The conclusion after several investigations became clearer: to remove all effects of radiation from outside (earth, sun and universe) our system, it is necessary to measure the background for 24, 48 or 72 hours. In the same way, the samples have to be measures for 24, 48 or 72 hours to be safe to be purified the measurement (equality of day and night measurement). It is also possible to not use the background of the winter in summer. Depend to the energy of radionuclide we seek, it is clear that the most important steps of a gamma spectrometry measurement are the preparation of the sample and the calibration of the detector

National Presentation - Repatriation of sealed high activity source (SHARS) from Cameroon

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