Shielding Requirements Research Articles

Estimating the Shielding Requirements for a Newly Planned PET-CT Facility

Positron Emission Tomography (PET) combined with Computed Tomography (CT) is a powerful and very sensitive diagnostic tool that integrates functional and anatomical imaging into one combined scanning system. Positron emission tomography is based on the characteristic way in which positrons annihilate by combining with an electron. This process usually results in the emission of two 511 KeV photons which travel in opposite directions. These 511 KeV annihilation photons are much higher energy than other diagnostic radiations. Because of this high energy of the annihilation radiation, shielding requirements for a PET facility are different from most of the other diagnostic imaging facilities and it’s a very important consideration in the design of a PET or PET-CT imaging facility. As a result, significant shielding may be required in floors and ceilings as well as adjacent walls in a PET-CT facility. In this work we present the estimation of the shielding requirements for a newly planned PET-CT facility. Shielding calculations of adjacent walls were presented for both controlled and uncontrolled areas. Formulas were used to calculate the shielding materials following the basic AAPM (American Association of Physicists in Medicine) guidelines. This mathematical analysis of the shielding estimation is very important for a newly planned PET-CT facility. Adequate safe planning with vendor, facility architect and a qualified medical physicist are essential to make a cost effective and safe design while maintaining radiation safety standards with regulatory limits.
 Bangladesh J. Nuclear Med. 21(2): 102-107, July 2018

Transient response predication of nickel coated carbon fiber composite subjected to high altitude electromagnetic pulse

Metal materials are gradually replaced by fiber reinforced composite and the electromagnetic shielding capacity needs to be quantitatively evaluated, especially for intense electromagnetic pulse. Equivalent permittivity and permeability of fiber composite are calculated by multi-scale model based on finite element method, which results are close to theoretical values. Transverse electrical conductivity is obtained by contact network model and numerical results are in agreement with experimental data. The propagation characteristics of high altitude electromagnetic pulse penetration through fiber reinforced composite is investigated by numerical method based on electromagnetic wave theory in time domain. It is found that fiber orientation and stacking sequence can affect the shielding capacity of fiber composite. Composite laminates stacked alternately at 45 degrees satisfy the shielding requirement when high altitude electromagnetic pulse incidence at any angle. Nickel-coated carbon fiber composite laminate show better shielding than that of carbon fiber composite laminate.

Initial integration concept of the DEMO lower horizontal port

The realisation of a Demonstration Fusion Power Reactor (DEMO) is the remaining and crucial step towards the exploitation of fusion power. This is the shared view of Europe and the nations engaged in the construction of ITER. DEMO will follow ITER and must be capable to produce several hundred MW of net electricity as well as operating with a closed fuel-cycle. The DEMO machine has three main entrance levels to the plasma chamber. According to the current DEMO reference configuration the vacuum vessel has 16 vertical upper, horizontal equatorial, and horizontal lower ports, respectively.This article introduces the initial integration concept of the lower port. The concept considers the external space constraints e.g. due to the supports of the vacuum vessel, the neutron shielding requirements of the superconducting coils, and the main lower port functions. These include hosting a number of tokamak systems, in particular the torus vacuum pump and feeding pipes of in-vessel components, and to allow for divertor remote maintenance. The size and position of the lower port are constrained by the adjacent toroidal and poloidal field coils. At the same time the lower port drives the layout of the cryostat and the tokamak building.

Analytical Prediction of High-Velocity Impact Resistance of Plane and Curved Thin Composite Targets

Resistance to ballistic impacts is a key requirement for an effective use of composite structures in many engineering applications. In fact, when fiber-reinforced composites are subject to high-velocity impacts (HVI), they are likely to suffer complete perforation, eventually causing the dramatic failure of the overall system. Moreover, elements with shielding requirements often feature nontrivial curvatures, so that their use in many engineering applications must undergo a precise evaluation of the effects of curvature on their resistance to impacts. The objective of the present work is to present an analytic formulation to model high-velocity impacts on both plane and curved (cylindrical and spherical) thin woven fabric composite targets, relying on the geometric and the elastodynamic characteristics of the target.

Conceptual design of the HCCB blanket system integration for CFETR

In a fusion power plant, the integration of the blanket system in the design progress is of vital importance to address the fundamental function of sufficient tritium production, reliable nuclear heat extraction, and permanent components protection due to the complex assembling systems into the tokamak vessel. Some progress activities of the blanket system design and integration for China fusion engineering test reactor (CFETR) are developed. The integration work involves the design of the breeding blanket, back plate support, shielding blanket, and supporting structures. To guarantee normal operation of the reactor, the design and arrangement of cooling pipes are very critical. The layout of the complex cooling pipes inside the blanket system is designed and integrated. Interfaces between main connecting components are designed for blanket system integration. In this work, the U-shaped HCCB blanket is utilized as the adaption breeding module into the integrated blanket system. The three-dimensional (3D) neutronic analyses verified that the integrated design of the blanket system could well meet the requirements of tritium self-sufficiency and neutron shielding. The nuclear heat generation of main components in the reactor is obtained with a nuclear energy multiplication factor of 1.35. The primary principle during the integration is to accommodate all the components allocated on the vacuum vessel. This work is the preliminary integration design and validation for CFETR blanket system, and further detailed design will be performed around these obtained references for better fusion feasibility and performance in the next-step design stage.

A Methodology for Optimizing the Management of Spent Fuel of Nuclear Power Plants Using Dry Storage Casks

The management of spent nuclear fuel assemblies of nuclear reactors is a priority subject among member states of the International Atomic Energy Agency. For the majority of these countries, the destination of such fuel assemblies is a decision that is yet to be made and the “wait-and-see” policy is thus adopted by them. In this case, the irradiated fuel is stored in on-site spent fuel pools until the power plant is decommissioned or, when there is no more racking space in the pool, they are stored in intermediate storage facilities, which can be another pool or dry storage systems, until the final decision is made. The objective of this study is to propose a methodology that, using optimization algorithms, determines the ideal time for removal of the fuel assemblies from the spent fuel pool and to place them into dry casks for intermediate storage. In this scenario, the methodology allows for the optimal dimensioning of the designed spent fuel pools and the casks’ characteristics, thus reducing the final costs for purchasing new Nuclear Power Plants (NPP), as the size and safety features of the pool could be reduced and dry casks, that would be needed anyway after the decommissioning of the plant, could be purchased with optimal costs. To demonstrate the steps involved in the proposed methodology, an example is given, one which uses the Monte Carlo N-Particle code (MCNP) to calculate the shielding requirements for a simplified model of a concrete dry cask. From the given example, it is possible to see that, using real-life data, the proposed methodology can become a valuable tool to help making nuclear energy a more attractive choice costwise.

Radiation Oncology-External-Beam Radiation Therapy.

The application of structural shielding design techniques and goals as outlined in the National Council on Radiation Protection and Measurements Report 151, Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities (2005), continues to be the basis for treatment vault design in 2018 with some updated information. Treatment techniques have changed significantly with the dominant usage of intensity-modulated radiation therapy techniques today based on concurrent imaging. Some of the developments in linear accelerator technology over the past 15 y include flattening filter-free modes, which enable higher instantaneous dose rates; three-dimensional conformal radiation therapy resulting in potentially higher workloads since healthy tissue is spared; improved intensity-modulated radiation therapy treatment delivery systems with lower monitor units per centigray delivered than traditional step and shoot intensity-modulated radiation therapy; stereotactic body radiation therapy with higher treatment fractions and increased workloads; and increased use of stereotactic radiosurgery with conventional linear accelerators as well as robotic arm-mounted linear accelerators with higher treatment fractions. These new treatment units also incorporate multiple-energy x-ray beams (2 to 5 MV typical), which require a significant change in the specification of a workload to be used in each vault. As the equipment in radiation oncology departments has evolved to state-of-the-art modalities, the requirements for adequate radiation shielding for these modalities has become more rigorous. Architectural designs no longer depend on standard maze design rectangular rooms. Innovative layouts and utilization of multiple layers of shielding materials allow much greater flexibility in room designs. Maze-less rooms with direct-shielded doors are part of these challenging designs and are very common today. Use of multiple-density concrete blocks allows quicker construction of vaults and requires less space for the equivalent shielding provided. Combinations of high-density or normal-density concrete, steel, and lead are used in designs to make optimum use of available space and cost. Additional shielding needed at the edges of these single- or bi-parting sliding doors as well as baffle designs for heating, ventilation, and air conditioning systems and communication cable penetrations require detailed calculations. Examples of these designs will be given in this presentation. Consideration of the radiation levels around the planned vault must also include adjacent multistory buildings.

Optimal configuration of a tokamak reactor with constraints moderately extrapolated from the ITER model

In this study, the optimal radial build and system parameters of a tokamak reactor were found by utilizing a coupled analysis of tokamak systems and neutron transport. Neutron impacts on shielding and tritium breeding capability were self-consistently incorporated, together with plasma physics and tokamak engineering constraints, which were moderately extrapolated from the ITER model. In a low-aspect ratio tokamak reactor, the minimum major radius to produce a desired fusion power was mainly determined by the shielding requirements, while in a normal aspect ratio tokamak reactor, it was determined not only by the requirements on the shielding, but also by the requirements on the tritium breeding and the magnetic flux density at the toroidal field (TF) coil. As the aspect ratio increased, the minimum major radius and the system size decreased as long as the tritium self-sufficiency was satisfied with only an outboard blanket, but they began to increase as the inboard blanket thickness increased to meet the requirements for tritium self-sufficiency and the TF coil bore radius increased to meet the requirements for the magnetic flux density at the TF coil. The fusion energy gain Q increased as the fusion power increased and as the confinement characteristics improved.

Equatorial electron cyclotron port plug neutronic analyses for the EU DEMO

Within the Power Plant Physics and Technology (PPPT) programme in the EUROfusion Consortium design activities are currently in progress for the development of a DEMOnstration Fusion Power Plant (DEMO). The design of the machine and the integration of in-vessel components require neutronics analyses to verify the tritium self-sufficiency, the shielding requirements, and the structural integrity of its components. In particular, the effect of penetrations in the blanket and in equatorial port plug introduced by the electron cyclotron (EC) heating system, namely due to openings for the antenna waveguides (WGs), were analyzed. In this study three-dimensional MCNP calculations were conducted for the pre-conceptual designs of the EC port plugs and shielding optimization were performed in order to ensure that the DEMO design limits are not exceeded.Two configurations of the EC heating system were performed using a DEMO Water Cooled Lithium Lead (WCLL) with integrated EC configurations model of a half of the sector (i.e. 10° model) and relevant ones repeated with a full sector model (i.e. 20° model) to test the reliability of results. Additionally, the effect of the radiation on the WGs closest to the plasma was analyzed as well as the impact on Tritium Breeding Ratio (TBR).

One-dimensional nuclear design analyses of the SST-2

Steady State Tokamak-2 (SST-2) will be an intermediate fusion machine before Indian DEMOnstration power reactor (DEMO) development to realise the reactor technologies. It is designed for fusion gain $$Q=5$$ and fusion power in the range of 100–300 MW. Nuclear design analyses of SST-2 machine have been carried out to support the conceptual design work. Analyses have been carried out for two breeding blanket concepts: Indian lead–lithium ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB). The analyses assess the tritium production and radiation shielding capability of the machine referring to the engineering design parameters. In this study, one-dimensional radiation transport calculations have been performed to assess the SST-2 nuclear responses for 1 full power year (FPY) operation. Nuclear responses such as tritium breeding ratio (TBR), various radiation loads to toroidal field (TF) coil have been calculated to obtain the radial build-up of SST-2 capable of breeding tritium and satisfying the shielding requirements. The assessment has been made using the ANISEN code and FENDL 2.1 cross-section library. It is observed that the TBRs with LLCB and HCCB blankets are 0.85 and 0.94, respectively. Shielding calculations confirm that the radial build is sufficient to protect the superconducting TF coils for 1 FPY.

Radiation Shielding Analysis and Design of tof PET-CT Facility at Institute of Nuclear Medical Physics under Bangladesh Atomic Energy Commission, Bangladesh

High resolution (4 mm) tof PET-CT (positron emission tomography-computed tomography) from Philips of model Ingenuity TF is newly installed at Institute of Nuclear Medical Physics (INMP). 128 slice CT component incorporated with PET provides comparatively lower dose than the 511 keV annihilation photons associated with positron decay from PET scan. So, for designing shielding in our PET-CT facility, only 511 keV annihilation photons energy has been considered. The main objective of this paper is to show what measures have been taken to protect patients, occupational workers as well as environment from PET-CT radiation hazard through a cost effective design that satisfy the national regulatory demand. In this paper, AAPM (American Associations of Physicists in Medicine) Task group 108 analysis for PET and PET-CT shielding requirements is followed for our PET-CT facility shielding design. From theoretical calculation as shielding requirement, 1.1 cm Pb thickness or, 13 cm concrete thicknesses are found. Practically, all walls and ceiling are of 30.48 cm (1 foot) thick made of concrete with density 2.35 gcm-3 for more safety. As x-ray from CT is not taken into account for shielding analysis, Bangladesh Atomic Energy Commission (BAEC) conducted an extensive radiation survey at controlled, supervised and public area for CT. The report that is found meets the national regulatory requirements.

Stretchable liquid metal electromagnetic interference shielding coating materials with superior effectiveness

The GIN amalgam can easily decorate various substrates and provide sufficient shielding protect. Stretchable film decorated by GIN can satisfy military shielding requirements (>60 dB) even at strain up to 75%.

Radiation Shielding for Helical Tomotherapy Vault Design.

Purpose:The purpose of the present study is to carry out radiation shielding calculations to find out adequate thicknesses of protective barriers such as walls and ceiling based on minimum space required to house helical tomotherapy unit. This study also aim to derive expression for use factor and estimation of patient workload for tomotherapy facility for optimizing radiation shielding requirements.Materials and Methods:The basic definitions and formulae given in NCRP/IAEA reports were referred and modified for tomotherapy machine to calculate optimized shielding thicknesses requirements. Workload is estimated based on observations of patient treatments on tomotherapy machine and analysis of their treatment plan data. A mathematical expression is derived for calculating use factor in terms of beam divergence angle at source corresponding to field length, angle of source rotation about isocenter, and distance of primary barrier from isocenter. Radiation shielding requirement of protective barriers such as walls and ceiling of helical tomotherapy vault is calculated based on minimum room dimensions as specified by the manufacturer, permissible dose limit (s), and values of optimizing parameters such as workload, use factor etc. for tomotherapy machine.Results:Using derived mathematical expression for use factor in this study, it was found that value of use factor varies with distance of primary barrier from isocenter and its value was found to be 0.093 for given minimum room dimensions. Radiation shielding requirements for protective barriers (walls/ceiling, etc.) were arrived and reported in this paper.Conclusions:A typical helical tomotherapy vault design is proposed based on the calculated shielding thicknesses of protective barriers. Further, it is also concluded that tomotherapy machine can be installed in a vault designed for 6 MV conventional linear accelerator with minor modification.

Radiation shielding requirements for the full power operation of the linear IFMIF prototype accelerator (LIPAc) at Rokkasho

The detailed shielding analysis has been performed in order to find a configuration of the radiation shielding in the accelerator building necessary for the operation of the Linear IFMIF Prototype Accelerator (LIPAc), which is under construction in Rokkasho, Japan, at the full power of 1.125 MW to satisfy the radiation dose limit prescribed by Japanese regulations. The final design of the beam dump constructed by CIEMAT, Spain was fully taken into consideration. The impact of all the features potentially causing neutron streaming were investigated and the optimization of local additional shieldings were conducted. As the result, the most important one was found to be the air ventilation duct shield, which completely covers the ducts and creates an additional labyrinth. By considering the additional shielding configuration determined in the present study, the legal criteria at the controlled area boundary can be completely satisfied.

Dependency of the capture of field emitted electron on the phase velocity of a high-frequency accelerating structure

Surface electric fields within high gradient linear accelerators can exceed 200 MV/m and lead to field emitted (FE) electrons entering the structure. When the accelerating field conditions permit, these FE electrons can become captured in the RF fields and be transported through the accelerating structure as a dark current. Understanding the capture and transport of these FE currents in high frequency linear accelerators, and at accelerating gradients well above the capture threshold, is important for the operation of CERN’s X-band test stands and other high gradient linear accelerators. Such dark current leads to a background radiation, which dictates shielding requirements and can damage adjacent instrumentation, as well as a background current within the structure, which can affect beam diagnostics and in the most extreme cases can cause transverse kicks on bunches. The capture of field emitted electrons is described analytically in a one dimensional approximation and is then evaluated numerically for a test structure geometry. A particular focus for the analysis is how the interaction varies with phase velocity. We demonstrate how the phase velocity varies with respect to the nominal driver frequency and structure operational temperature. Measurements on the X-band test stands at CERN demonstrate that the capture increases 12%–28% for a 1 MHz increase in the driver frequency. A three dimensional RF and particle simulation found a similar order of magnitude result for a 1 MHz increase corrborating the measurements.

A study regarding the optimal radial build of a low-aspect-ratio tokamak fusion system

System parameters and the optimal radial build of a low-aspect-ratio (LAR) tokamak fusion system were found through the coupled analysis of a tokamak system and neutron transport. In a configuration with only an outboard breeding blanket, the minimum major radius to produce a given fusion power was determined by the shielding requirements and the magnetic field at the toroidal field (TF) coil. With a confinement enhancement factor H = 1.4, Q > 10 was possible for fusion power greater than 800 MW with an aspect ratio of A = 1.5; however, Q > 10 was possible for fusion power greater than 1800 MW with an aspect ratio of A = 2.0. The outboard radial build was determined by the tritium breeding and shielding requirements. The tritium breeding capability of blanket concepts proposed for testing in the International Thermonuclear Experimental Reactor (ITER) was evaluated by varying the outboard blanket thickness and the degree of lithium-6 (Li-6) enrichment. Cases with a smaller aspect ratio exhibited better performance since the number of fusion neutrons that contributed to tritium breeding were larger than the case with a larger aspect ratio. Among the blanket concepts, a helium (He)-cooled solid breeder (HCSB) concept showed the best tritium breeding capability and thus allowed for a smaller system size.

Modeling and Simulation for Coolant Activity and Dose Rate Estimation in a Pool-Type Nuclear Research Reactor

Coolant activity in the primary coolant system and reactor pool in an open pool-type research reactor is very important in view of operational and radiological safety considerations. As pool water acts as the shielding medium to minimize the radiation dose at the top of the reactor pool, an estimation of the activities of radionuclides in pool water is essential to establish the safety of the operating personnel and researchers working at the reactor pool top. A system is provided to create a hot water layer (HWL) at the top of the pool by supplying water at a temperature more than that of the pool water so that the dose rate at the reactor pool top can be minimized. This HWL system helps in breaking the natural convection current of reactor pool water by maintaining a higher temperature at this layer so that high-density pool water below this layer cannot replace this low-density HWL. Therefore, pool water that is comparatively more radioactive will not be able to cross this HWL by convection. Hence, diffusion will be the only mechanism by which radioactivity can reach the pool top. So eliminating the convection current keeps the activity at this topmost layer of the pool at a minimum value. The estimation of the activity of the radioactive nuclides is required to assess the radiation field at different locations in the primary coolant loop for designing proper shielding requirements of the system. The radionuclides of interest are the activation products of aluminum (24Na, 27Mg, 28Al) and 41Ar. In this technical note, a transient code is presented for estimating the activity of radioactive nuclides in the coolant loop and reactor pool of a nuclear research reactor. The reduced activity level at the pool top is estimated considering the presence of the HWL at the top of the pool. It is observed that purification flow plays a major role on the activity level of radioactive nuclides in reactor pool water. The variation of the activity dose rate at the reactor pool top with purification flow is also discussed.

Monte Carlo simulation of neutron dose equivalent by photoneutron production inside the primary barriers of a radiotherapy vault

PurposeTo evaluate the neutron dose equivalent produced by photoneutrons inside the primary barriers of a radiotherapy vault. MethodsMonte Carlo simulations were performed for investigating the production of photoneutrons as well as neutron shielding requirements. Two photon beams of 15 and 18 MV struck sheets of steel and lead, and the neutron doses were calculated at the isocenter (Piso) and at a distance of 50 cm from the inside wall (Pwall) while delivering 1 Gy to the patient. The proper thicknesses of borated polyethylene (BPE) and concrete were simulated to reduce neutron contamination. ResultsWhen the primary barrier consisted of a concrete alone, the neutron doses at Piso were 0.5 μSv/Gy and 12.8 μSv/Gy for 15- and 18-MV, respectively. At Pwall, the neutron doses were 15.8 μSv/Gy and 318.4 μSv/Gy for 15- and 18-MV, respectively. When 15 MV photons interacted with metal sheets, the neutron doses were 0.4–22.2 μSv/Gy at Piso and 15.8–812.5 μSv/Gy at Pwall, depending on the thickness and material of the metal sheets and neutron shielding. In the case of 18 MV photons with the same configuration, the neutron doses were 0.9–59.5 μSv/Gy and 73.9–5006.1 μSv/Gy for Piso and Pwall, respectively. The neutron dose delivered to the patient was reduced to the level of the dose delivered with a concrete barrier by including a 10-cm-thick BPE for each beam. ConclusionsWhen the primary barrier shielding is designed with a metal sheet inside for high energy, proper neutron shielding should be constructed to avoid undesirable photoneutron dose.

Suppressing Electromagnetic Radiation by Trapping Ferrite Nanoparticles and Carbon Nanotubes in Hierarchical Nanoporous Structures Designed by Crystallization‐Induced Phase Separation

AbstractPolymer blends are at the forefront of research especially in the field of Electromagnetic Interference (EMI) shielding because of their versatile properties such as ease of processability, economic viability and high strength to weight ratio. Herein, we have attempted to design lightweight blend composites consisting of multiwalled carbon nanotubes (MWNTs) and nickel ferrite (NiFe2O4) nanoparticles. A unique approach was adopted here to prepare ultra‐thin (500 μm), flexible, lightweight composite membranes wherein hierarchical nanoporous structures were initially developed by crystallization induced phase separation in a classical upper critical solution temperature (UCST) pair Polyvinylidene fluoride/poly methyl methacrylate (PVDF/PMMA) and subsequently etching out the PMMA phase. In the next step, functional nanoparticles were trapped in the pores by facile vacuum filtration approach. This unique approach led to the fabrication of nanoporous composite membranes which otherwise is difficult to process using conventional techniques. The composite membranes show high magnetic permeability and high electrical conductivity; the two key requirements for effective shielding of electromagnetic (EM) radiation. A significant improvement in shielding effectiveness (SE) was achieved using these token composite membranes. For instance, porous PVDF composite membranes containing 3 wt % MWNTs (with a thickness of 500 μm) showed an SE of 8 dB which enhanced significantly to 27 dB for composite membranes wherein NiFe2O4 is trapped in the pores. More interestingly, the mechanism of shielding was driven by absorption (nearly 80%) through synergistic properties of interconnected MWNTs and NiFe2O4 nanoparticles.

Mitigation strategies against radiation-induced background for space astronomy missions

The Advanced Telescope for High ENergy Astrophysics (ATHENA) mission is a major upcoming space-based X-ray observatory due to be launched in 2028 by ESA, with the purpose of mapping the early universe and observing black holes. Background radiation is expected to constitute a large fraction of the total system noise in the Wide Field Imager (WFI) instrument on ATHENA, and designing an effective system to reduce the background radiation impacting the WFI will be crucial for maximising its sensitivity. Significant background sources are expected to include high energy protons, X-ray fluorescence lines, 'knock-on' electrons and Compton electrons. Due to the variety of the different background sources, multiple shielding methods may be required to achieve maximum sensitivity in the WFI. These techniques may also be of great interest for use in future space-based X-ray experiments. Simulations have been developed to model the effect of a graded-Z shield on the X-ray fluorescence background. In addition the effect of a 90nm optical blocking filter on the secondary electron background has been investigated and shown to modify the requirements of any secondary electron shielding that is to be used.