Sealed Sources Research Articles

Sites selection for a potential radioactive waste repository in Peninsular Malaysia: GIS-based weight linear combination and multi-criteria decision-making analysis

This paper aims to determine the suitable places for a radioactive waste repository in Peninsular Malaysia to implement a permanent disposal facility for disposal sealed radioactive sources (DSRS). The sole Malaysia interim waste storage that stores a large proportion of DSRS was expected to be occupied by 2025 fully. Currently, the nation’s licensing board promoted the borehole disposal facility to resolve the full capacity dilemma. The potential sites were determined by the geographical information system (GIS) based Weight Linear Combination (WLC) analysis, integrating the improved Analysis Hierarchical Process (improved AHP) and the technique for order of preference by similarity to ideal solution (TOPSIS) approaches. The sensitivity analysis was applied to identify the main contributing factors for corresponding Peninsular Malaysia selection. Ten parametric criteria and three exclusion criteria are applied to assess the potential areas and screen out the unfavorable regions, respectively. Results indicated the final composite map presents 435 areas of very high suitability class with 17.5% coverage from the total suitability class. The top 5 largest candidates with high potentiality are Jerantut & Ulu Lepar, Bandar Tun Razak, Pertang & Simpang, Pertang, Tanjung Ipoh, and Labis, where come from three provinces: Pahang, Negeri Sembilan and Johor. By TOPSIS model, the ranks for these five candidates are Jerantut & Ulu Lepar, Bandar Tun Razak, Tanjung Ipoh, Labis and Simpang Pertang. The model revealed that the suitability ranks is proportional to its area sizes in the approximate values of criteria. The three main contributing factors in Peninsular Malaysia are hydrogeology, seismicity and lithology, which are strongly recommended to involve in other corresponding site selection project.

Cesium immobilization in porous silica and [formula omitted]Cs self-heating simulations

The interest to immobilize 137Cs can pursue different aims. When it is separated from spent nuclear fuels, it is possible to reduce the thermal loading of the host glass and increase the waste loading capacity of the matrix. Also, the separated 137Cs can be used to produce radioactive sealed sources useful in industrial and medical applications. If reprocessing is not envisaged after the generation of electrical energy, the spent nuclear fuel is the most important radioactive waste generated in the nuclear fuel cycle. In this case, aqueous solutions obtained from the chemical dissolution of spent nuclear fuels are high-level wastes. Short-lived fission products, such as 90Sr and 137Cs, constitute the main source of heat generation from beta decay, causing self-heating of the glassy matrix. It is well known that vitrification is the main used technology, capable to immobilize these wastes during hundreds of years generating a waste form with proved structural stability, thermal shock resistance, and high chemical durability. In this article we present in an analogous way the immobilization of an specific amount of stable cesium in a porous Si-rich glass matrix through adsorption and sintering. Sintering requires lower temperature than that required in immobilization by melting. The leaching behavior of the waste form obtained was studied from the procedure described by the MCC-1 test method. Considering 137Cs only for simulations, this work also includes the thermal evolution calculation of a simulated silica glass block loaded with 2 wt.% of 137Cs.

Progress and Challenges of Radioactive Waste Management in Thailand

In 2016, the Thai government established a new Act, the Nuclear Energy for Peace Act. As a result, various regulations, requirements, and guidelines regarding radioactive waste (RW) management were established under the Act. The waste management activities conducted by the waste operator were carefully reviewed and demonstrated regulatory compliance. The operator can deal with the waste that currently exists. However, to maintain continuity for the safe and sustainable management of the waste, it is necessary to overcome various challenges, such as the decommissioning of the disused facilities that will be released from regulatory control. The contaminated metals require special attention, because they are increasing rapidly. The use of melting treatment technology and the monitoring of the radioactive contamination in materials might be a solution to reduce the quantities of these metals. The presence of cesium-137 (Cs-137) contaminated dust from steel production factories is a concern due to the large volume. No decision on the management of the dust has been made. However, researchers are now carrying out experiments to determine methods to deal with it. In addition, the operator should consider important issues, such as the development of an additional storage facility to support the decommissioning of waste and the dust waste, and the reuse and recycling of disused sealed radioactive sources (DSRS). This paper will discuss the aspects of waste disposal, the strategic plans that deal with the challenges, and naturally occurring radioactive materials (NORMs).

Surface degradation of laminated metal 226Ra foils used in school sealed sources in the UK

In the few years leading up to this research, CLEAPSS noticed a small but steadily increasing number of calls from UK schools regarding a red-brown discolouration on the surface of the foil of their radium source. There were no reports of this type of discolouration on foils of other radionuclides. CLEAPSS and the University of Liverpool collaborated to investigate the nature and cause of this discolouration and the likelihood that the foils were becoming unsafe. The evidence indicates that the discolouration is principally caused by some combination of silicon, sulfur and possibly lead from within the foil diffusing into the face layer. There is no indication currently that the face layers are fragmenting on these foils, but the longer-term integrity of the discoloured foils now becomes questionable. Given the age of the foils and the radiotoxicity of radium, the recommendation from this research is that discoloured foils should be taken out of service and disposed.

A dual-functional micro-focus X-ray source based on carbon nanotube field emission

A micro-focus X-ray source with miniature pressure sensor was developed based on the multi-walled carbon nanotube (MWNT) field emission. The field emitters for both X-ray cathode and the pressure sensor were synthesized by direct chemical vapor deposition growth of MWNTs on catalytic substrates. MWNTs exhibited excellent field emission properties with the threshold field at 10 mA/cm2 of 2.30 V/μm, and the high current density of 10 A/cm2 could be reached at the field of 4.36 V/μm. The field emission electron gun was constructed with the aid of electrostatic simulations on the electron beam trajectories and the focal spot size (FSS) on the target with the demagnification factor of 5. The practical objects, including the semiconductor chip and the ceramic heating element, were tested by the X-ray source with the spatial resolution of less than 35 μm, and the fracture of the ceramic element inside metal case could be clearly detected. The pressure sensor monitored the in-situ pressures inside the source during the processing and after the sealing-off. The pressures in the sealed sources were generally in 10−4 Pa level and could rise to 10−2 Pa level during the operation.

New Level of high-resolution Computed Tomography with a 450kV Meso Focus x-ray Source

Computed Tomography (CT) production applications of modern components are increasingly faced with the challenge of achieving a very high resolution even with high densities and a comparatively large material thickness. These components in the aerospace, automotive and electronics industries often consist of 3D-printed metals such as iron, copper, nickel or alloys with resolution requirements rising higher than 10 lp/mm (50 μm). With Computed Tomography and the new MesoFocus x-ray source at 450 kV, YXLON uses the innovative Comet technology, which is made for industrial applications where resolution and penetration of dense material-objects matter. CT with the MesoFocus sources eliminates the tradeoff everyone has to make today between sealed tube solutions and open microfocus tubes – sealed tubes are robust and stable and made for production environments, but offer only limited resolution, while open microfocus tubes offer high resolution, but are not made for production environments and 24/7 operation. YXLON’s new CT solutions allow hassle-free operation, while they bring lab resolution to the production floor. The excellent stability of the sealed MesoFocus x-ray source enables long exposure times that are an enormous advantage for CT applications and are a perfect fit for inspection in aerospace, automotive and electronics industries. The excellent resolution down to 30 μm makes the CT solution the key enabler for advanced manufacturing technologies such as additive manufacturing. We will present CT results for 3D-printed iron and copper parts as well as for finest mechanics surrounded by gold with sub40μm resolution. Duplex wire IQI and ASTM E1695 results prove the image chain resolution.

Prompt gamma-ray methods for industrial process evaluation: A simulation study

Abstract Radioisotope applications in industrial process inspection and evaluation using gamma-ray emitters provide otherwise unavailable information. Offering alternative gamma-ray sources can support the technology by complementing sources’ availability and radiation safety. This work proposes to replace gamma-ray from radioisotopes with prompt gamma-ray from the interaction of neutrons with stable isotopes injected into the industrial process or with the structural material of the industrial process equipment. Monte Carlo N-Particle Transport Code (MCNP5) was used to simulate the irradiation of two-phase flow pipes by 252Cf neutron source. Two simulations were run for each pipe, with and without mixing the liquid phase with the stable isotope 157Gd. The detected gamma-ray spectra were analysed, and images of the two phases inside the pipes were produced. The images were compared to images obtained from simulations of gamma transmission measurement using 60Co. Furthermore, results for prompt gamma computed tomography (CT) were presented and discussed. The studies’ outcomes indicate the potential of prompt gamma-ray to carry out the sealed sources applications of gamma transmission measurements and imaging.

Monte Carlo optimum management of 241Am/Be disused sealed radioactive sources

The optimum encapsulation of 241Am/Be disused sealed radioactive sources (DSRS) based on PHITS Monte Carlo simulations for their long-term storage in Cameroon was performed. The country capacity for the management of disused neutron sources was also evaluated and showed that a Am1 P60 capsule is sufficient for the total available inventoried 241Am/Be DSRSs. The effective dose rate was computed in the enclosures of the DSRS container, which will be temporarily stored in the centralized radioactive waste facility. The obtained results were in agreement with the ALARA principle for the exposure rate optimization and the obtained exposure dose rates were found to be 1.830 μSv/h (horizontal calculation) and 0.137 μSv/h (vertical computation) which values are lower than the 2.5 μSv/h acceptable limit for the public area. The dose profile for 241Am/Be source obtained, the neutron flux, and gamma generated from neutron absorption showed agreement with the research hypothesis. The Monte Carlo assessment achieved in the present research will be useful for dismantling and preparing the waste package for long-term storage.

Ionizing radiation metrology at the service of health: quality control of radiopharmaceutical dose calibrators in nuclear medicine unit -accurarcy, reproducibility and linearity tests-

Ensuring the effective use of a dose calibrator used for the practice of nuclear medicine, and the injection of the correct dose to the patient is the subject of a long traceable calibration chain, which serves to establish the relationship between the values of the activity indicated by the dose calibrator and the activity corresponding to ISO 2919 certified sealed sources whose activity is known with an uncertainty of less than 1 %. Our calibration program consists in carrying out various quality control on the CRC 55TR activimeter, namely the daily control which is carried out in the absence of any radioactive source, then the fidelity control carried out by a source of 37Cs and a source of 60Co, then the reproducibility check performed with a 137Cs source, and finally the linearity check performed with a 99mTc source. In our case the results obtained from the specific tests used to verify the proper functioning of the Capintec CRC 55tR activimeter in a nuclear medicine service in Morocco, shows an agreement with international standards so that the measurement of activities is done with significant results, in addition, to guarantee the metrological quality of the measurements delivered, Our laboratory registers in an accreditation process according to ISO 17025.

HEXITEC 2 × 2 tiled hard X-ray spectroscopic imaging detector system

HEXITEC is a spectroscopic imaging X-ray detector technology developed at the STFC Rutherford Appleton Laboratory for X-ray and γ-ray spectroscopic imaging applications. Each module has 80 × 80 pixels on a 250 μm pixel pitch, and has been implemented successfully in a number of applications. This paper presents the HEXITEC 2 × 2 detector system, a tiled array of 4 HEXITEC modules read out simultaneously, which provides an active area of 16 cm2. Systems have been produced using 1 mm thick Cadmium Telluride (CdTe) and 2 mm thick Cadmium Zinc Telluride (CdZnTe) sensor material. In this paper the system and data processing methods are presented, and the performance of the systems are evaluated. The detectors were energy calibrated using an 241Am sealed source. Three types of charge sharing correction were applied to the data-charge sharing addition (CSA), charge sharing discrimination (CSD), and energy curve correction (ECC) which compensates for energy lost in the inter-pixel region. ECC recovers an additional 34 % of counts in the 59.5 keV peak in CdTe compared to the use of CSD; an important improvement for photon-starved applications. Due to the high frame rate of the camera system (6.3 kHz) an additional End of Frame (EOF) correction was also applied to 6.0 % of events to correct for signals that were readout whilst the signal was still forming. After correction, both detector materials were found to have excellent spectroscopic performance with a mean energy resolution (FWHM) of 1.17 keV and 1.16 keV for CdZnTe and CdTe respectively. These results successfully demonstrate the ability to construct tiled arrays of HEXITEC modules to provide larger imaging areas.

Development and pre-clinical test of a phosphorous-32 containing polyetheretherketone foil aiming at urethral stricture prevention by low-dose-rate brachytherapy.

PurposePrimary or recurrent urethral stenoses are common clinical problems. The aim of this feasibility study was development and application test of a novel radioactive catheter for potential use in low-dose-rate (LDR) brachytherapy to prevent urethral stenosis.Material and methodsA beta radiation emitting phosphorous-32 (32P)-foil was developed and integrated in indwelling catheter, which is usually inserted after surgical intervention in the urethra. Activity and dose distribution were measured after neutron activation of the phosphorus-31 (31P)-foil component with scintillation techniques as well as radiochromic films, and compared to MC simulations. Practicability and radiation safety of this new applicator were tested on male rabbits, which, before catheter application, had undergone a new method of stricture induction by radial laser light irradiation.ResultsMeasured dose distributions of catheters with 32P-foils were found to be in good agreement with MC simulations. Wash-out tests confirmed the radioactive catheter setup to be inside permitted limits of a sealed source. Strictures could be induced by laser light in all animals, surgically treated by slitting and irradiation with a 7-day dose of 15 Gy or 30 Gy, without adverse effects of radiation during a 4-week observation time.ConclusionsThis proof-of-concept study presents a safe and technically simple approach to LDR brachytherapy intending to prevent urethral stenosis. The prescribed irradiation dose can be administered in a reproducible and precise manner using a urethral catheter as a carrier of beta radiating foil. The study duration of 63 days was too short to make statistically significant stricture-related conclusions. This approach may be considered for similar problems in other hollow organs and in cases regarding superficial hyper-proliferation.

Application of radioactive particle tracking and an artificial neural network to calculating the flow rate in a two-phase (oil–water) stratified flow regime

A multiphase flow is defined as the transport of two or more fluids with different properties flowing together inside a pipeline. After offshore oil production, it is necessary to control the amount of transported fluids based on flow rate measurements. Therefore, in this study, we developed a simulation method for predicting the volume fraction and calculating the superficial velocity for a two-phase flow based on radioactive particle tracking, which involves using a sealed radiation source inside the pipeline in order to obtain volume fraction measurements. The test section for the multiphase flow comprised oil and saltwater under a stratified flow regime, with a polyvinyl chloride pipe, four NaI(Tl) detectors, and a137Cs radioactive particle that emitted gamma-rays at 662 keV. Simulations were conducted using the MCNP6 code, which is a mathematical code based on the Monte Carlo method. Volume fraction predictions were obtained using a multilayer perceptron neural network with a backpropagation algorithm. The novel feature of this method is the combination of radioactive particle tracking with an artificial neural network in order to predict volume fractions in multiphase flows. The results showed that 91.65% of the predicted patterns were within 5% of the relative error. In addition, the time delay was determined using the cross-correlation function to obtain the superficial velocity in three different volume fractions, which allowed each phase flow rate to be calculated in these cases.

National nuclear forensics libraries: a case study on benefits and possibilities for identification of sealed radioactive sources

A National Nuclear Forensic Library (NNFL) is a useful nuclear forensics tool which consists of information and subject matter expertise on radioactive and nuclear (RN) materials produced, used or stored within a State. If RN material is found out of regulatory control the NNFL can be used as part of a nuclear forensics investigation to help identify whether or not the material is consistent with a country’s national holdings. In previous work, a number of signatures which can be useful to identify sealed sources of 241Am were investigated. To validate the measurement results, an official query concerning information about two of the previously investigated 241Am sources was sent to the United States Department of State, the international point-of-contact (POC) for the U.S. NNFL. The aim of this work is to show how data obtained in a characterization of a radioactive source can be used in conjunction with an NNFL to investigate the history of a source out of regulatory control.

Contamination Resulting From a Broken 125I Seed During a Brachytherapy Procedure.

Brachytherapy programs within radiation therapy departments are subject to stringent radiation safety requirements in order to ensure the safety of the staff and patients. Training programs often include brachytherapy-specific radiation safety training modules that address the specific risks associated with radioactive sources, emergency procedures, and regulatory requirements specific to the use of radioisotopes. Unlike other uses of radioactive materials, brachytherapy uses sealed sources and therefore under routine operations does not encounter radioactive contaminants. This article presents an unusual clinical situation in which an 125I brachytherapy seed was damaged during routine clinical workflow, resulting in radioactive contamination within the clinical environment. Decisions made at the time of the incident resulted in contamination that spread beyond the initial location. The incident highlighted a shortcoming of the radiation safety program in preparing staff for the possibility of having to deal with unsealed radioactivity. Brachytherapy programs would be strengthened by including training specific to radioactive contamination in their emergency training to equip staff to respond to unexpected damage to the sealed sources.

241Am/Be source optimum geometry for DSRS management-based Monte Carlo simulations

The geometry form related to 241Am/Be spontaneous neutron sources has been investigated to evaluate the most optimizing geometry for disused sealed radioactive source (DSRS) disposal. Nine source geometries were assessed: point, disk, sphere 1, sphere 2, sphere 3, cylinder 1, cylinder 2, rectangle 1, and rectangle 2. The most radiological optimizing geometries were found to be the disk, followed by the point source, sphere 1, and cylinder 1, while the rectangle or parallelepiped 2 source forms were the worst cases. Neutron and photon generated fluxes were computed, and the parallelepiped 2 source geometry generated the lower value of flux inversely to the effective dose rate computed in the decision-making area, where it was found to be the highest value. The obtained results were in agreement with the as low as reasonably achievable principle for the exposure rate optimization. The obtained exposure dose rates were found to be <∼2 µSv/h (maximum value on horizontal calculation) and 1.75 µSv/h (maximum value on vertical computation), which are lower than the 2.5 µSv/h acceptable limit for the public area. Whenever possible, the DSRSs should be pressed into a disk form (or a cylindrical form with a small height) before embedding them into the Am1 P60 capsule for disposal purposes. The dose profile for the 241Am/Be source obtained, the neutron flux, and the gamma generated from neutron absorption showed agreement with the expected experimental physical data. The supported data contribute to postulate that Monte Carlo methods are effective computational tools that can be used to select the most effective radioactive waste form for disposal purposes.

Preparation of 134Cs sealed source based on WO3 adsorbent for radiometric calibration purposes

Cesium-134 artylon sealed source based on inorganic adsorbent (as a core) was easily prepared. The sealed source was in the form of a disk constituted of two halves; base and top. Cesium-134 was adsorbed by equilibration of 134CsCl solution with WO3 inorganic adsorbent at pH 4 for 24 h at 25±1°C. Then, the liquid phase was decanted and checked for the remaining activity. Thereafter, the 134Cs-loaded WO3 was dried at 100 °C for 24 h. A definite weight the 134Cs-loaded WO3 adsorbent was placed in the central cavity of the base half-disk. Then the two half-disks were assembled using a commercial epoxy adhesive. The activity values of the sealed source were 7.81×10-2 and 7.70×10-2 Ci, calculated by using the absolute efficiency curve and DETEFF software were found to be closest at a 5 cm source-detector distance; respectively. After radiometric standardization and encapsulation, the sealed source was checked radiometrically before release to the local market for calibration purposes of radiation measurement and identification instruments.

Energy calibration of the GEMPix in the energy range of 6 keV to 2 MeV

The GEMPix is a small gaseous detector with a highly pixelated readout, consisting of a drift region, three Gas Electron Multipliers (GEMs) for signal amplification, and four Timepix ASICs with 55 μm pixel pitch and a total of 262,144 pixels (512×512 pixels). A continuous flow of a gas mixture (here propane-based tissue equivalent gas) is supplied externally at a rate of 5 L/h. By placing a sealed 241Am source outside of the detector and varying the source-detector distance, the residual energy of alpha particles entering the sensitive volume of the GEMPix through a thin Mylar window is varied. An alpha spectrometry measurement was performed to determine the emission spectrum of the source, and this spectrum was then used as an input to a FLUKA Monte Carlo simulation to obtain the residual energy of the alpha particles. Thus, a calibration curve from 5.9 keV (from an 55Fe source) up to more than 2 MeV was obtained, which is needed for future applications of the detector, in particular for microdosimetry.

Radioactive Iodine-125 in Tumor Therapy: Advances and Future Directions.

Radioactive iodine-125 (I-125) is the most widely used radioactive sealed source for interstitial permanent brachytherapy (BT). BT has the exceptional ability to deliver extremely high doses that external beam radiotherapy (EBRT) could never achieve within treated lesions, with the added benefit that doses drop off rapidly outside the target lesion by minimizing the exposure of uninvolved surrounding normal tissue. Spurred by multiple biological and technological advances, BT application has experienced substantial alteration over the past few decades. The procedure of I-125 radioactive seed implantation evolved from ultrasound guidance to computed tomography guidance. Compellingly, the creative introduction of 3D-printed individual templates, BT treatment planning systems, and artificial intelligence navigator systems remarkably increased the accuracy of I-125 BT and individualized I-125 ablative radiotherapy. Of note, utilizing I-125 to treat carcinoma in hollow cavity organs was enabled by the utility of self-expandable metal stents (SEMSs). Initially, I-125 BT was only used in the treatment of rare tumors. However, an increasing number of clinical trials upheld the efficacy and safety of I-125 BT in almost all tumors. Therefore, this study aims to summarize the recent advances of I-125 BT in cancer therapy, which cover experimental research to clinical investigations, including the development of novel techniques. This review also raises unanswered questions that may prompt future clinical trials and experimental work.

Borosilicate glass 60Co high dose rate brachytherapy thermoluminescence dosimetry

Brachytherapy is commonly used in treatment of cervical, prostate, breast and skin cancers, also for oral cancers, typically via the application of sealed radioactive sources that are inserted within or alongside the area to be treated. A particular aim of the various brachytherapy techniques is to accurately transfer to the targeted tumour the largest possible dose, at the same time minimizing dose to the surrounding normal tissue, including organs at risk. The dose fall-off with distance from the sources is steep, the dose gradient representing a prime factor in determining the dose distribution, also representing a challenge to the conduct of measurements around sources. Amorphous borosilicate glass (B2O3) in the form of microscope cover slips is recognized to offer a practicable system for such thermoluminescence dosimetry (TLD), providing for high-spatial resolution (down to < 1 mm), wide dynamic dose range, good reproducibility and reusability, minimal fading, resistance to water and low cost. Herein, investigation is made of the proposed dosimeter using a 1.25 MeV High Dose Rate (HDR) 60Co brachytherapy source, characterizing dose response, sensitivity, linearity index and fading. Analysis of the TL glow curves were obtained using the Tmax-Tstop method and first-order kinetics using GlowFit software, detailing the frequency factors and activation energy.

Applicability of polymer and composite inner linings in containers for borehole disposal of sealed radioactive sources − A simulation-based study of radiation effects

Borehole disposal concept is a widely applicable, standardized solution for disposal of sealed radioactive sources, developed jointly by IAEA and South African NECSA with the goal of offering a safe and secure disposal solution, based on readily accessible gear and materials. Previous research has considered use of polymer and composite materials in place of standard materials such as concrete and steel in various phases of radioactive waste management. Use of alternatives such as these can be advantageous in environments that have adverse effects on commonly utilized materials, in terms of chemical aggressiveness or radiation damage. The present study examines the suitability of polymers and polymer-based composites for potential use as inner lining in borehole disposal containers, from the perspective of radiation effects. Investigated materials, not envisioned by the original concept, include HDPE, PS, PEEK, PEI and PSU, as well as composite materials having one of these polymers as a matrix reinforced by boron fibers. Such containers would be simpler and safer to manipulate, owing to the lower mass of these materials. Applicability of the investigated materials is analyzed by using Monte Carlo simulations to determine the energy deposited or absorbed dose in various container linings. These results are then used for calculating the overpressure within the container, caused by radiation-induced (i.e. radiolytic) gas generation, as well as the changes in polymer molecular weights which may influence their mechanical characteristics. Radioactive sources used in simulation models, positioned inside disposal containers of known standard dimensions, include Co-60 and Cs-137 gamma sources and the americium-beryllium (AmBe) neutron source. The obtained results indicate that the examined materials do not produce any significant increase in container gas pressure or experience notable changes in molecular weights, and are hence a viable option for linings in borehole disposal containers.