Radiation Shielding

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THE SOLUTION IS RIDICULOUSLY SIMPLE.  ALL IDIOTS CAN DO IT, EVEN YOU - DOING SO IS VERY GOOD FOR THE SOUL  -  AND IT PROTECTS AGAINST RADIATION

Background: Biological shielding of nuclear reactors has always been a great concern and decreasing the complexity and expense of these installations is of great interest. In this study, we used datolite and galena (DaGa) minerals for production of a high performance heavy concrete. Materials and Methods: Datolite and galena minerals which can be found in many parts of Iran were used in the concrete mix design. To measure the gamma radiation attenuation of the DaGa concrete samples, they were exposed to both narrow and wide beams of gamma rays emitted from a cobalt-60 radiotherapy unit. An Am-Be neutron source was used for assessing the shielding properties of the samples against neutrons. To test the compression strengths, both types of concrete mixes (DaGa and ordinary concrete) were investigated. Results: The concrete samples had a density of 4420-4650 kg/m 3 compared to that of ordinary concrete (2300-2500 kg/m 3 ) or barite highdensity concrete (up to 3500 kg/m3). The measured ...

A study of the shielding properties of a metamaterial slab is presented. The metamaterial nature of the slab arises from the negative values of its effective scalar dielectric permittivity in the considered lowfrequency region of operation. Such an effective permittivity is the result of a homogenization process of an actual periodic structure which consists of a finite number of periodic layers of thin conducting cylinders embedded in a dielectric matrix, screening an E-polarized plane wave. When the incident plane wave is arbitrarily polarized, a more sophisticated model for the effective permittivity has to be employed which takes into account anisotropy and spatial dispersion. Comparisons with conventional planar metal screens show that the proposed structure can be designed to be advantageous in terms of low density and weight savings.

Introduction
The purpose of this study was to evaluate personal shields in radiation departments of hospitals affiliated to Mashhad University of Medical Sciences.
Materials and Methods
First, the information related to 109 personal shields was recorded and evaluated by imaging equipment. Afterwards, the equivalent lead thickness (ELT) of 62 personal shields was assessed, using dosimeter and standard lead layers at 100 kVp.
Results
In this study, 109 personal shields were assessed in terms of tears, holes and cracks. The results showed that 18 shields were damaged. Moreover, ELT was evaluated in 62 shields. As the results indicated, ELT was unacceptable in 8 personal shields and lower than expected in 9 shields.
Conclusion
According to the results, 16.5% of personal shields had defects (tears, holes and cracks) and 13% of them were unacceptable in terms of ELT and needed to be replaced. Therefore, regular quality control of personal shields and evaluation of new shields are necessary at any radiation department.
Keywords: Occupational Exposure; Radiography; Shield

Energy consumption and efficient use are important factors of thermal systems. This work indicates numerical studies on surface radiation and the effects of natural convection on the heat transfer and flow field between two united cylinders united, using a radiation shield between them. This study reveals the physical and engineering effects of the radiation shield on the internal radiation of two concentric cylinders closed at a certain temperature. It consists of two concentric cylinders with a hotter inner cylinder and a cooler outer cylinder. A 2 mm thick radiation shield (made of aluminum oxide plate) is inserted between the cylinders. Radiation protection has been found to reduce radiation from the inner surface of the outer cylinder to the outer surface of the inner cylinder. Almost 28% of the hypothermia is observed due to decreased radiation by the radiation shield.

The measurement of outdoor environmental and climatic variables is needed for many applications such as precision agriculture, environmental pollution monitoring, and the study of ecosystems. Some sensors deployed for these purposes such as temperature, relative humidity, atmospheric pressure, and carbon dioxide sensors require protection from climate factors to avoid bias. Radiation shields hold and protect sensors to avoid this bias, but commercial systems are limited, often expensive, and difficult to implement in low-cost contexts or large deployments for collaborative sensing. To overcome these challenges, this work presents an open source, easily adapted and customized design of a radiation shield. The device can be fabricated with inexpensive off-the-shelf parts and 3-D printed components and can be adapted to protect and isolate different types of sensors. Two material approaches are tested here: polylactic acid (PLA), the most common 3-D printing filament, and acrylonitrile styrene acrylate (ASA), which is known to offer better resistance against UV radiation, greater hardness, and generally higher resistance to degradation. To validate the designs, the two prototypes were installed on a custom outdoor meteorological system and temperature and humidity measurements were made in several locations for one month and compared against a proprietary system and a system with no shield. The 3-D printed materials were also both tested multiple times for one month for UV stability of their mechanical properties, their optical transmission and deformation under outdoor high-heat conditions. The results showed that ASA is the preferred material for this design and that the open source radiation shield could match the performance of proprietary systems. The open source system can be constructed for about nine US dollars, which enables mass development of flexible weather stations for monitoring needed in smart agriculture.

A new technique was introduced in this study to utilize the prompt gamma rays, that generated from the fission process, in gamma irradiation experiments. The method de- pends on utilizing the difference of the attenuation effect of water on neutrons from that of gamma rays
The core of 22 MW power is submerged in the water of open pool type reactor and the theoretical and experimental studies have illustrated that the influence of water in attenuating the prompt neutrons is higher than the prompt gamma rays.
Based on this phenomena, the dose rate distributions for the prompt neutrons and the prompt gamma rays were studied in the water volume that surrounding the core during reactor operation. Then, the proper position, that achieves the considerable gamma dose rate as well as the minimum neutron dose rate, could be determined. An irradiation device was developed to contain the samples during the gamma irradiation process. The irradiation device would be located in the proper position that was chosen depending on the previous conditions. MAVRIC code was used to determine the dose rate distributions in the water surrounding the reactor core.

Describes, with examples, two methods for selecting and adjusting proportions for normal weight concrete, both with and without chemical admixtures pozzolanic, and slag materials. One method is based on an estimated weight of the concrete per unit volume; the other is based on calculations of the absolute volume occupied by the concrete ingredients. The procedures take into consideeration the requirements for placeability, consistency, strength, and durability. Example calculations are shown for both methods, including adjustments based on the cheracteristics of the first trial batch.

Nuclear medicine is a medical specialization that uses radioactive materials injected into the body to diagnose and treat human diseases. The use of different radionuclides and high amounts of radioactive materials makes it necessary for the facilities where these procedures are conducted to evaluate the corresponding shielding to comply with the design dose limits of a facility and avoid radiological accidents as recommended and accepted in international publications, like the International Commission on Radiological Protection (ICRP) and the National Council on Radiation Protection and Measurements (NCRP). This work compares two methods to calculate the shielding necessary to guarantee que las medicine service zones be safe from ionizing radiations. The first method consists in calculating the transmission factor B to find the thickness of the material necessary to protect the zone of interest, this factor is calculated by bearing in mind the occupancy factors, workloads, use factor, and the design objective dose limit. Upon obtaining the transmission factor B, half value layer (HVL) or tenth value layer (TVL) tables are used for each construction material, obtaining the thickness of the material. The other method is the calculation of is the calculation of rates of exposure through the air Kerma rate constant, then the XCOM databases are used, which were developed by the National Institute of Standards and Technology (NITS) to obtain the attenuation coefficient, used in the law of exponential attenuation; finally, the necessary thickness of the material is obtained to reach the design objective. Finally, the principal differences between both methods are shown and an analysis is performed of the shielding optimization, seeking to set criteria to make recommendations to nuclear medicine services on optimal shielding.. Resumen La medicina nuclear es una especialidad médica que utiliza materiales radioactivos inyectados en el cuerpo para diagnóstico y tratamiento de enfermedades humanas. El uso de diferentes radionúclidos y altas cantidades de materiales radiactivos hace necesario que las instalaciones donde se realicen estos procedimientos evalúen los blindajes correspondientes para cumplir los límites de dosis de diseño de una instalación evitando así accidentes radiológicos recomendados y aceptados en las publicaciones internacionales como el ICRP (International Commision on Radiological Protection) y el NCRP (National Council on Radiation Protection and Measurments). En este trabajo se comparan dos métodos para el cálculo de los blindajes necesarios para garantizar que las zonas del servicio de medicina sean seguras a las radiaciones ionizantes. El primer método consiste en calcular el factor de transmisión B para hallar el espesor del material necesario para proteger la zona de interés, este factor se calcula teniendo en cuenta los factores de ocupación, cargas de trabajo, factor de uso y el límite de dosis objetivo de diseño. Una vez obtenido el factor de transmisión B se usan las tablas de HVL (Half Value Layer) o TVL (Tenth Value Layer) para cada material de construcción obteniéndose el espesor del material. El otro método es el cálculo de las tasas de exposición por medio de la constante de la tasa de Kerma en Aire, luego se usan las bases de datos (XCOM) desarrolladas por NITS (National Institute of Standards and Technology) para obtener el coeficiente de atenuación que son utilizados en la ley exponencial de atenuación; finalmente, se obtiene el espesor de material necesario para alcanzar el objetivo de diseño. Finalmente, se muestran las principales diferencias entre los dos métodos y se hace un análisis de la optimización de los blindajes buscando tener criterios para hacer recomendaciones a los servicios de medicina nuclear sobre blindajes óptimos.

A new computer code named as SHIELD DESIGN has been developed in Visual Basic Version 6.0 for precise estimation of shielding thickness in gamma ray shielding designs. In the conventional forward calculations using point kernel method, the sources and geometry are specified along with the shield thickness to obtain dose rates at desired locations. But in this code, a new computational technique based on inverse calculation and iterative procedures have been employed to estimate the shield thickness using the dose rate criteria as input information. This new technique enables quicker and easier way of shield design optimization by computing shield thickness required in a single run whereas in the forward shield design optimization method, dose rate computations need to be carried out for various shield thicknesses. The shielding thickness arrived with this code provides a first hand information to the designer to carry out forward calculations for accurate optimization. Also, the code is very useful to regulators for quicker verification of shield calculations submitted by the designers for approval. Further, various options available in this code extend its flexibility and user friendliness. This paper briefly presents the development and validation of the code, SHIELD DESIGN. The validation has been carried out by performing analyses of the gamma shielding design for some standard problems. The results were compared with those of renowned forward calculations, experimental measurements, NCRP-51 benchmarks and other gamma radiation shielding utilities like RAD-PRO and satisfactory agreement was observed. These validation tests demonstrate the precise estimation of shield thickness.

Shielding of transmission line is a vital role in the power system because most of the power outage is occurred due to non-scientific shielding, machine problem, manual error and environmental conditions. Hence, these issues cause disturbance in electrical flow which in turn effects the living beings, transformer, machine etc. thus it is important to analyze the fault present in transmission line which helps out to remove electrical problem within short period and to re-establish the system on minimum interference. This interruption of power system can be brought up to an end by detecting and classifying the different electrical faults on transmission line with the help of sensors and then intimates the electricity board about fault and its location through a microcontroller, GSM module and serial communication device.

GAMOS is a software system for GEANT4-based simulation. It comprises a framework, a set of components providing functionality to simulation applications on top of the GEANT4 toolkit, and a collection of readymade applications. It allows to perform GEANT4-based simulations using a scripting language, without requiring the writing of C þ þ code. Moreover, GAMOS design allows the extension of the existing functionality through user-supplied Cþ þ classes. The main characteristics of GAMOS and its embedded functionality are described.

This paper presents analysis of lightning arrester energy due to back flashover and shielding failure for the purpose of installing transmission line arrester (TLA). This study simulated, analysed and validated with the installed lightning arresters on the 132kV transmission line in Malaysia by using PSCAD/EMTDC software. The arrester is based on IEEE frequency dependent model. The simulation result shows that the lightning arrester energy obtained through simulation is less than the rating installed at site © 2012 IEEE.

http://ieeexplore.ieee.org/xpls/abs_all.jsp?arnumber=6307344&tag=1

Radioisotope technology is applicable in medicine, agriculture, hydrology and various industries. Its versatility is increasing with advent of knowledge, research and development. The radioisotope technology may be termed as of zero carbon emission. The green quotient of the technology further improves as it consumes some of the by- products of nuclear power technology. Application of radioisotope technology in all aspects of life requires even spread of radioisotope facilities. The equipments being used for various application of this technology require very precise and controlled application. The scientific, technical and engineering domain of this technology is still evolving.
Constructions of such facilities are different from conventional buildings due to its environmental, health, commercial and strategic importance. Such constructions involve micrometer level accuracy and almost zero tolerances. These issues are to be addressed in totality covering planning, material selection, construction methodology with a focus on final utility.
Leak tight, corrosion free with minimal contamination probability construction of radio active zones is an area of concern, for which special lining methodology and site specific quality assurance measures have to be developed and adopted. Construction of hot cell and radio-activity related areas are a key activity in such facilities. The main construction material, in normal structures, is reinforced concrete and the same has to be utilised for making technology affordable. The needs for accurately positioning many special embedded parts and through pipes at various locations, with adequate radiation shielding provisions are challenges to be addressed.     
The construction methodologies, techniques, special formwork systems etc. need special attention. The site specific measures and innovations to cope up the challenges encountered such as heat of hydration during concreting, and void-less dense concrete around fixtures also merit attention. The strict quality control measures adopted for a fail safe construction output and after construction serviceability of utility is of vital importance.
This paper briefly describes a recently constructed project in the light of prevailing guidelines for such structures. Planning, safety, quality and construction methodologies and innovations used at site with reference to better adoptability of the project has also been highlighted.

The shielding performance of a planar metamaterial wire-medium (WM) screen under plane-wave illumination is studied. The screen consists of a finite number of periodic layers of thin conducting cylinders embedded in a dielectric matrix; in the lowfrequency limit and for waves with the electric field polarized along the wire direction, such a screen can be modeled as a homogeneous slab with a plasma-like dispersive permittivity. In particular, well below the plasma frequency, the effective relative permittivity is negative and very large in absolute value, giving rise to high values of the shielding effectiveness. A comparison with a conventional planar metal screen is reported, showing how the proposed structure can be designed to be advantageous in terms of low density and weight saving. In order to be effective against arbitrarily polarized waves, the original structure is modified, adding a second WM screen with wires orthogonally oriented with respect to the first one. Numerical results are provided that illustrate the validity of the homogenized model in predicting field levels both near to and far from the screen and show the shielding performance attainable with the proposed metamaterial screen in both single-and double-layer configurations at normal incidence.

h i g h l i g h t s Barite and hematite replaced serpentine by 25 and 50% of its concrete volume. Unlike barite, hematite increased the porosity of serpentine concrete (A) Barite and hematite decreased the mechanical strength of concrete A. 25% hematite revealed the highest Rth for concrete A. 50% barite showed the highest c-ray attenuation properties for concrete A. a b s t r a c t To highlight the impact of barite and hematite addition on selected physical, mechanical, and radiation attenuation properties of serpentine concrete (A), different replacement ratios of barite and hematite aggregates (25, and 50% by volume) were used separately. The physico-mechanical properties of the prepared serpentine-barite (AB25, 50) and serpentine-hematite (AH25, 50) concrete mixes compared to those of the control (A) were evaluated. The radiation shielding properties of these concrete mixes against both c-ray and thermal neutrons were experimentally measured. The linear attenuation coefficient of c-ray, m, and thermal neutron attenuation coefficient, Ʃ th , were determined experimentally by the transmission method in a fine beam geometry setup using 60 Co c-ray source and a moderated PuBe source, respectively. The experimental results of c-rays were verified by the correlation with the theoretical data obtained by the WinXCom program. Besides, many c-ray attenuation parameters, such as a half-value layer, HVL, single-valued, and energy-dependent effective atomic number, Z eff , as well as effective electron density, N eff , were determined. It was found that the addition of barite and hematite aggregates to serpentine concrete increased its bulk density, and decreased its mechanical strength. Moreover, unlike barite, the hematite increased the water absorption and porosity of concrete. In conclusion , this study revealed that the prepared hematite-and barite-serpentine concrete mixes, except for AB50, have acceptable mechanical properties with enhancing radiation shielding performance.

Background: Polyethylene composites including boron can be used as an effective neutron shield. Our investigation focuses on manufacturing borated polyethylene nano-composite. The purpose of this study is to design a radiation shield for use in both neutron and gamma fields. Materials and Methods: Borated polyethylene shields containing 2%, and 5% weight percentage of Boron nano-particles were constructed and their neutron attenuation was compared with pure polyethylene. Polycarbonate films were used to find the attenuation of Am-Be neutrons after passing the shields. Mechanical properties of the shields were finally compared. Results: Mean  (± SD) number of α tracks induced by neutrons passing through the shields, were found to be 1.0488×103±128.98,  1.1972×103±289.56and 1.5340×103±206.52 for polyethylene with 5% by weight, polyethylene with 2% by weight boron nano-particles, and pure polyethylene, respectively. The neutron spectrum after each shield was also obtained by MCNP4C Monte Carlo simulations. On the other hand, borated polyethylene nano-composites showed higher tensile strength compared to that of pure polyethylene. Attenuation of neutrons measured in experiments and the result of MCNP simulation were in good agreement. Conclusion: A statistically significant difference was found between neutron attenuation by borated polyethylene nanocomposite made of 5% by weight boron and pure polyethylene. However, the difference between borated polyethylene nano-composite with 5% weight and 2% wt boron was not statistically significant.

The term  head and neck  cancer refers to a  group  of  biologically similar  cancers  originating  from  the  upper  aero  digestive  tract, including  the  lip,  oral  cavity  ,  nasal  cavity,  Para  nasal  sinuses, pharynx, and larynx. 90% of head and neck cancers are squamous cell carcinomas, originating from the mucosal lining (epithelium) of these regions. Radiation therapy  is the  most common form  of treatment along  with surgery and chemotherapy. Radiotherapy is the art of using ionizing radiation to destroy malignant cells while minimizing  damage  to  normal  tissue.  Radiotherapy  has  become astandard treatment  option for a wide range  of  malignancies, but adverse tissue reactions associated with the use of radiotherapy in the management of patients with head and neck cancer are painful and they diminish the quality of life. So a novel approach has been made to fabricate customized intraoral stents that can help prevent
the unnecessary irradiation of the surrounding normal tissues, thus reducing  the  severity  ofreactions.  However,  when  properly designed  and  used,  these  stents  are  effective  in  reducing  the treatment morbidity.

Three types of lead shielding assemblies were investigated to establish which provides the lowest background as preliminary work for determining the efficiency of a High Purity Germanium (HPGe) detector. The background continuum and peaks were examined using gamma spectroscopy with a coaxial and a planar HPGe detector. The 210Pb concentration in each lead sample taken from the assemblies was quantitatively determined. A French lead sample was found to have the lowest background in this relative comparison with a 210Pb concentration of 29.171.3 Bq kg-1.

The variation of radiation shielding parameters like linear and mass attenuation coefficients, half value layer, tenth value layer and mean free path with different soil samples of various chemical compositions and physical properties has been investigated using gamma ray spectrometer NaI(Tl) atthe gamma energies of 360, 662, 1280, 1330 KeV emitted from radioactive isotopes of 133 Ba, 137 Cs, 22 Na and 60 Co, respectively. For this work, soil samples were collected from different sites of Aurangabad, India, and then prepared using standard techniques. The study of attenuation coefficients of various materials has been an important part of research in radiation physics, chemistry, agriculture and human health. This study aims to investigate of radiation shielding properties for various soils of various components. For the attenuation coefficients, a significant variation is observed in low and high energy regions whereas there is no notable change in the intermediate region. The results have been discussed on the basis of obtained attenuation coefficients of different partial photon interaction processes. It is observed that the variation of obtained values of all parameters strongly depends on the photon energy; it decreases or increases due to chemical composition and density of the sample. The study has practical importance to know the nature of the soils in the construction and agricultural purposes. Generally, the radiation shielding parameters determined in this study can be used for determination of gamma emitters in any soil samples.

This work studies the design of the treatment room for an 18-MV linac to ensure radiation protection and safety of hospital staff and patients. The walls' thickness, the door, and the maze were designed according to the National Council on Radiation Protection and Measurements Report 151 recommendations. The results of this work are contrasted with the Monte Carlo calculations performed with the MCNP5 code where dose equivalents due to neutrons and neutron spectra estimated at different points inside and outside the radiotherapy room verify that the shielding thicknesses obtained are enough to reduce the dose level permitted by Mexican regulation.

This paper presents analysis of lightning arrester energy due to back flashover and shielding failure for the purpose of installing transmission line arrester (TLA). This study simulated, analysed and validated with the installed lightning arresters on the 132kV transmission line in Malaysia by using PSCAD/EMTDC software. The arrester is based on IEEE frequency dependent model. The simulation result shows that the lightning arrester energy obtained through simulation is less than the rating installed at site.

To modify some properties of serpentine-based concretes, serpentine aggregate (S) was partially replaced with either hematite aggregate (H) or barite aggregate (B) at levels of 25% and 50%, by volume. To evaluate the impact of these aggregates, selected physical properties of the prepared serpentine-barite (AB 25 , 50) and serpentine-hematite (AH 25, 50) concretes were measured and compared to those of the control serpentine concrete (A). In addition, the fast neutrons and gamma-rays shielding properties of these concretes in comparison with those of the control (A) were experimentally measured using a 239 Pu-Be source and stilbene detector. These shielding properties were determined by measuring the neutron and gamma-ray attenuation coefficients (Σ R ðE n Þ and μðE γ Þ, respectively), as well as parameters of relaxation length (λ) and half value length (HVL) at different energies. It was found that incorporation of B and H had a negative impact on the physical properties of the serpentine-based concretes except for density. On the other hand, this incorporation improved the attenuation properties of serpentine-based concretes. This improvement was more effective in the case of using B rather than H in the equivalent ratios. Finally, 50% B was the best ratio in enhancing the shielding of fast neutrons and gamma-rays.

Theoretical perspective discussions about negatively charged graphene slabs and their radiation properties are presented. It is shown that additional electrons in graphene, injected in a charging process, can enhance the value of the total cross section of photon radiation interactions, especially for X-rays and lower energies. A detailed discussion about the photoelectric effect on electrons trapped in quantum wells in graphene supplements the investigations on light carbon radiation shields. It is concluded that hundreds of parallel graphene slabs would give a significant contribution to modern radiation protection.

The effectiveness of the shielding against ionizing radiation in controlled and supervised areas of four (4) selected imaging rooms of Mulago hospital was determined. Scattered radiation transmitted to the operator's console, leakage through the walls and doors to the patient waiting areas was also determined. The effective doses to occupational workers were also determined. The availability and effectiveness of the lead aprons and other protective gears was also investigated. Thermoluminiscent Dosimeters (TLDs) were installed at selected points for a period of four (4) weeks. Radiation leakages to the members of the public were measured using a survey meter, scattered X-ray radiation to staff was measured using TLD badges. The mean scattered radiation in the imaging rooms varied from 1.19 mSv/month in the Computed Tomography (CT) room to 0.38 mSv/month from the Casualty Center (CC). The effective doses to occupational staff were highest in Room 4 (R4) (plain radiography) of 6.8 mSv/yr and lowest in CC at 1.4 mSv/yr. Radiation leakages through selected doors were found to be 18.1 mSv/hr at the Uganda Cancer Institute (UCI). Though there was some leakage, the available shielding was found to be generally effective.

Mammography x-rays uses to demonstrate different lesions among women by using ionizing radiation. This irradiation exposes the patient to slight high radiation amounts. This study conducted in amongst mammography patients in Khartoum, Sudan. The sample magnitude was 100 patients whose different types of diseases. All patients were underwent mammography. The factors were documented in this study were patient's characteristics and radiation exposure related factors. The mean MGD values calculated were 3.6 + 0.29 (p<0.05). Those results were suggestively more than the standard dose determined by international agencies (IAEA, NCRP and ACR).

The evaluation of the shielding effectiveness of metallic enclosures with apertures containing metallic objects and excited by arbitrary sources is the main topic of this work. The shielding effectiveness is computed through an integral formulation based on the Method of Moments which makes use of a fast computation of the enclosure Green functions. Absorbing artificial materials are also designed and proposed to damp the resonant fields excited inside the enclosure in order to improve the shielding performance.

The shielding effectiveness of artificial periodic screens is investigated with reference to high-frequency highimpedance near fields produced by arbitrarily oriented electric dipoles. The screens are formed by metallic objects of arbitrary shape displaced in a 2-D lattice. The problem is first studied through a full-wave approach, using the array scanning method in conjunction with a periodic method of moments in the spatial domain. Next, the shielding problem is solved analytically in some characteristic electromagnetic compatibility configurations through the use of approximate low-frequency homogeneous models together with a classical analysis in the spectral domain. Finally, the solutions are compared with those deriving from the use of the so-called transmission-line approximation. The provided results show the suitability of the analytical approaches in dealing with finite sources different from standard plane-wave excitations and give a useful tool for the design of periodic shields.

This study suggests an alternative path to transport the cobalt device after irradiation in open-pool-type reactor from the reactor core to the cobalt cell. The cobalt transport process in current path depends on using heavy shielded cask, 1500 Kg weight, under water surface of spent fuel storage pool which may cause a missile impact of spent fuel accident (Falling of heavy load into spent fuel storage pool). To avoid using the shielded cask, an alternative path is suggested depending on transporting the cobalt device through the hot cells of the reactor; the testing and the cobalt cells. The expected radiation dose rate rise associated with the alternative transport path must be evaluated at positions which the worker will be found during transport process. SCALE/MAVRIC sequence code was used to determine the dose rate levels along the suggested transport path; over spent fuel storage pool water surface, and around testing and cobalt cells. The calculated results show rise in radiation dose levels more than the permissible dose rate limits around testing cell. Suggested procedures are presented in this study to decrease the received dose by worker found around testing cell during transport process.

The effects of apertures with finite thickness on the shielding effectiveness of rectangular metallic enclosures is investigated through an approximate but efficient formulation based on the method of moments. As expected, the analysis shows that the finite aperture thickness can help to increase the shielding effectiveness of the structure, especially in the highfrequency region. Comparisons with rigorous full-wave results confirm the validity of the proposed technique.

High density concrete is extensively used for efficient radia on a enua on in radiotherapy rooms and nuclear reactors. Over the past eight years, some efficient galena-based concrete samples for shielding X or gamma rays was produced. The goal of this study was to produce a novel high density concrete against neutron and photon radia ons using tourmaline and galena.

IFMIF-DONES (International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source) is currently being developed in the frame of the EUROfusion Early Neutron Source work package (WPENS). It will be an installation for fusion material testing, that will generate a flux of neutrons of 10 18 m −2 s −1 with a broad peak at 14 MeV by Li(d,xn) nuclear reactions thanks to a deuteron beam colliding on a liquid Li flow. The accelerator system is in charge of providing such high energy deuterons in order to produce the neutron flux expected. The objective of the Beam Dump, part of the High Energy Beam Transport Line (HEBT), is to stop the pulsed beam at low duty cycle during DONES accelerator commissioning and start-up phases. The present work explains the radiological design of the beam dump shielding and two different configuration approaches for the materialization of the design. The radiological design considers maintenance and operation, and it was done together with the building walls dimensioning so that the combined radiation attenuation by the local shield and the building leads to dose rates in the different rooms that satisfy the requirements. Activation of the materials in the HEBT line, originated by the leakage of neutrons through the beam dump entrance is evaluated and an ad-hoc solution is proposed for its minimization. Regarding the mechanical design, in the first configuration, the shielding is split into two halves horizontally, the upper-half requiring external lifting capabilities for its commission and maintenance. The second approach consists in a vertical splitting into two halves, which are self-moveable, avoiding the needs of external lifting capabilities for the remote handling of the shielding.

Mixed neutron and gamma radiations require different shielding materials as their interaction with materials is different. Composites were developed in order to combine the shielding capabilities of different materials. However, their homogeneity is difficult to be assured which can lead to pinholes where radiation can penetrate. To avoid this problem, several materials arranged in layers can be used to shield against mixed radiations. Since the multilayer shielding can be made from any material in many configurations, the ant colony optimization (ACO) is a promising method because it deals with combinatorial optimization problems. The candidate materials are HDPE, boron, cadmium, gadolinium, tungsten, bismuth, and iron. Preliminary MCNP simulations were done to observe the effect of arrangements, thicknesses, and types of materials on the radiation spectrum. It was found that: (1) the final layer should be made of high density material, (2) an increase beyond certain thicknesses did not result in a significant increase in attenuation, and (3) there should be an optimum combination of material that can effectively shield against both neutrons and gamma rays.

Regulatory entities require that for any radiation facility the surrounding areas must be restricted unless the dose equivalent is less than 0.02 mSv in any one hour. Two Monte Carlo radiation transport simulation codes, MCNP5 and Mercurad, were used to design a facility to shield a 3.48 × 105 MBq 137Cs irradiator that meets these requirements. Simulations showed that the dose equivalent rates were below the legal limit for unrestricted access and the facility was constructed using available concrete block and student labor to minimize costs. To verify the accuracy of the Monte Carlo radiation transport codes, an ion chamber was used to characterize the facility. Ion chamber measurements in the actual, as-built irradiation facility showed that the Monte Carlo codes, MCNP5 and Mercurad, agreed by a factor of better than 6% and better than 11%, respectively.

Original Article BACKGROUND Environmental contamination and degradation at mining sites always result in waste rocks and waste such as tailing piles and can accumulate over the years of mining activities. Mining influenced water, including contaminated surface water, groundwater, and seepage from former mine openings, waste in the form of slurry that find their way into abandoned tantalite mines, waste sludge that was discharged into unlined lagoons, and aerial deposition of heavy metals and other contaminants from ore processing Introduction: Tantalite and other solid mineral deposits in Oke-Ogun occupy a very large landscape, and the possibilities of further exploration in new landscape and abandoning unproductive sites cannot be ruled out. These landscapes when abandoned become hideout for criminals and ritualistic activities which pose a great threat to the wellness of the people in the area and Nigerians at large. Objectives: This study attempts to investigate the radiological safety of these abandoned mines so as to convert them to economical valuable sites that can serve as a means of employment for population of the unemployed youth in the area and Nigeria at large. Materials and Methods: A total of seventeen soil samples were randomly selected. These samples were processed and analyzed for activity concentrations of 226 Ra, 232 Th, and 40 K using Hyper Pure Germanium Gamma Spectrometer. RESidual RADioactivity software version (6.5), a computer program developed by the Environmental Assessment Division of Argonne National Laboratory, was used for dose prediction for 70 years using the activity concentrations of soil samples from all the selected sites as its input parameter for the software. Predicted doses were estimated with respect to the three exposure pathways, namely external gamma irradiation from radionuclides in the soil samples, inhalation of dust particles, and inadvertent ingestion of soil. Results: The measured activity concentrations of radionuclides: 226 Ra for all the sites is in the range of (17-177) Bq/kg, 232 Th in the range of (2-107) Bq/kg, and 40 K in the range of (650-1667) Bq/kg. The predicted dose for all the sites is within the range of 0.05-0.35 mSv/y with Eluku mining site having values greater than the permissible limit of 0.25 mSv/y. The health implication of these values is that three of the sites (Gbedu, Sepenteri, and Komu) are safe for other usages. Conclusion: Any of these three mining sites may be used for recreational purposes, fishing and building of estates to mention but a few.

The aim of the present work is to develop a computational model of a compact pressurized water nuclear reactor (PWR) to investigate the use of innovative materials to enhance the biological shielding effectiveness. Two radiation transport codes were used: the first one – MCNP – for the PWR design and the GEM/EVENT to simulate (in a 1D slab) the behavior of several materials and shielding thickness on gamma and neutron radiation. Additionally MATLAB Optimization Toolbox was used to provide new geometric configurations of the slab aiming at reducing the volume and weight of the walls by means of a cost/objective function. It is demonstrated in the reactor model that the dose rate outside biological shielding has been reduced by one order of magnitude for the optimized model compared with the initial configuration. Volume and weight of the shielding walls were also reduced. The results indicated that one-dimensional deterministic code to reach an optimized geometry and test materials, combined with a three-dimensional model of a compact nuclear reactor in a stochastic code, is a fast and efficient procedure to test shielding performance and optimization before the experimental assessment. A major outcome of this research is that composite materials (ECOMASS 2150TU96) may replace (with advantages) traditional shielding materials without jeopardizing the nuclear power plant safety assurance.

─ The shielding problem consisting in the interaction between a dipole source and a Frequency-Selective Surface (FSS) is investigated. The Array Scanning Method (ASM) is adopted to take into account all the propagating and evanescent waves, which constitute the spectrum of the dipole and all the propagating and evanescent Floquet modes, which constitute the spectrum of the diffracted field by the FSS. The main differences with respect to the shielding of a conventional plane-wave source are pointed out, especially in terms of resonant frequencies, operating bandwidth and transmission levels.

Nuclear medicine technicians would receive unavoidable exposures during the preparation and administration of radiopharmaceuticals. Based on the staff dose monitoring, the dose reduction efficiencies of the radiation protection shields and the need to implement additional strategies to reduce the staff doses could be evaluated. In this study, medical staff doses during the preparation and administration of Tc-99 m, I-131, and Kr-81 radiopharmaceuticals were evaluated. The dose reduction efficiencies of the lead apron and thyroid shield were also investigated. GR-207 thermoluminescent dosimeter (TLD) chips were used for quantifying the medical staff doses. The occupational dose magnitudes were determined in five organs at risk including eye lens, thyroid, fingers, chest, and gonads. TLDs were located under and over the protective shields for evaluating the dose reduction efficiencies of the lead apron and thyroid shield. The occupational doses were normalized to the activities used in the working shifts. During preparation and injection of Tc-99 m radiopharmaceutical, the average annual doses were higher in the chest (4.49 mGy) and eye lenses (4 mGy). For I-131 radiopharmaceutical, the average annual doses of the point-finger (15.8 mGy) and eye lenses (1.23 mGy) were significantly higher than other organs. During the preparation and administration of Kr-81, the average annual doses of the point-finger (0.65 mGy) and chest (0.44 mGy) were higher. The significant dose reductions were achieved using the lead apron and thyroid shield. The radiation protection shields and minimum contact with the radioactive sources, including patients, are recommended to reduce the staff doses.