Sealed Sources Research Articles

THE PROBLEMS OF USING EXEMPTION ACTIVITY VALUES FOR REGULATING THE MANAGEMENT OF SEALED RADIONUCLIDE SOURCES OF GAMMA-RADIATION

The article focuses on the procedure for exemption of the sealed and unsealed radionuclide sources of gamma radiation from regulatory control. The contradictory nature of the existing set of exemption criteria has been noted, leading, in some cases, to paradoxical situations. It is shown that the exempt activity values determined in NRB-99/2009 and in the international basic safety standards of the IAEA are significantly overestimated (in comparison with the activity values of a point source creating the ambient dose equivalent rate of 1 μSv / h at a distance of 0.1 m) for a number of the most widely used gamma-emitting radionuclides, including 22 Na, 54 Mn, 75 Se, 15 2Eu and 154 Eu. It is proposed to revise the current values of exempt activity, bringing them in line with the dose rate criterion for the exempt of sealed radionuclide sources of gamma radiation, and to present them with one significant digit. The corrected values of exempt activity for seven selected radionuclides are proposed for further use in the process of revision of the national radiation safety standards.

Risk Analysis for Overexposure Measurements Due to Radiological Accidents Using Computer Code and the Lessons Learned “Meet Halfa as a Case Study, Egypt"

Large number of accidents that have occurred in three major practices, namely: industrial radiography, gamma irradiators, and radiotherapy. A radiological accident involves a sealed or unsealed radiation source leads to an uncontrolled release of ionizing radiation or radioactive materials into the environment, which may result in significant human exposure and/or material damage. A major accident occurred in greater Cairo Area during 2000 due to bad practice and human error. A radiological source (Ir-192 capsule with activity 31.5 Ci) was lost at the examination site near the pipe. Members of the public were plucked up. At least seven persons have exposed to overdoses, two of them died due to overexposure dose. In this study, the absorbed radiation doses and the associated hazards due to uncontrolled radioactive sources Ir-192 and Cs-137 are calculating using RESRAD‑Building computer code. An overview of the consequences and the obtained results are registered in lessons learned study would clarify the necessary that the regulatory authority to introduce measures capable to avoid the recurrence of similar accidents.

Performance of BLC-200 Cask under 9m Drop Test

BLC-200 cask is designed to carry 200 kCi Co60 sealed sources. The cask is made of stainless steel, lead and tungsten. Lead is used as primary shielding material, whereas tungsten is used in the inner periphery to make the cask compact and light. One of the main design features of BLC-200 is its inclined fins which not only act as a shock absorber during impact but also helps in dissipating heat during normal conditions of transport. The cask is designed as a Type B (U) package as per International Atomic Energy Agency (IAEA) & Atomic Energy Regulatory Board (AERB) regulations, according to which the cask should maintain its structural integrity when dropped onto an unyielding target from a height of 9 m. Numerical simulation with Explicit Finite Element Technique is used to study the performance of the cask under 9 m drop test. The paper brings out the methodology of analysis and performance of cask under 9 m drop test. Numerical simulations of the cask under various drop orientations are presented. Results such as deformation, g-loading, lead slump and bolt stresses induced are discussed.

The American College of Radiology and the American Brachytherapy Society practice parameter for the performance of low-dose-rate brachytherapy

Brachytherapy is the use of radionuclides to treat malignancies or benign conditions by means of a radiation source placed close to or into the tumor or treatment site. This practice parameter refers only to the use of radionuclide brachytherapy. Brachytherapy alone or combined with external beam therapy plays an important role in the management and treatment of patients with cancer. Low-dose-rate (LDR) brachytherapy has traditionally been used for treating prostate, head and neck, breast, cervical, and endometrial cancers as well as obstructive bile duct, esophageal, or bronchial lesions. It has been practiced for over a century with a variety of sources including radium-226, cesium-137, and, more recently, iridium- 192, iodine-125, and palladium-103. Low-dose-rate (LDR) brachytherapy can be given as interstitial, intracavitary, intraluminal, and/or plesiotherapy to a wide variety of treatment sites. This practice parameter addresses sealed sources as they are used for LDR brachytherapy. It is recognized that unsealed sources (e.g., yttrium-90) are also a form of LDR brachytherapy.

Collapsed cone dose calculations for heterogeneous tissues in brachytherapy using primary and scatter separation source data

Background and ObjectiveBrachytherapy is a form of radiation therapy using sealed radiation sources inserted within or in the vicinity of the tumor of, e.g., gynecological, prostate or head- and neck cancers. Accurate dose calculation is a crucial part of the treatment planning. Several reviews have called for clinical software with model-based algorithms that better take into account the effects of patient individual distribution of tissues, source-channel and shielding attenuation than the commonly employed TG-43 formalism which simply map homogeneous water dose distributions onto the patient. In this paper we give a comprehensive and thorough derivation of such an algorithm based on collapsed cone point-kernel superposition, and describe details of its implementation into a commercial treatment planning system for clinical use. MethodsA brachytherapy version of the collapsed-cone algorithm using analytical raytraces of the primary photon radiation followed by successive scattering dose calculation for once- and multiply scattered photons is described in detail, including derivation of the corresponding set of recursive equations for energy transport along cone axes/transport lines and the coupling to clinical source modeling. Specific implementation issues for setting up of the calculation grid, handling of intravoxel gradients and voxels partly containing non-patient applicator material are given. ResultsSample runs for two clinical cases are shown, one being a gynecological application with a tungsten-shielded applicator and one a breast implant. These two cases demonstrate the impact of improved dose calculation versus TG-43 formalism. ConclusionsUse of model-based dose calculation algorithms for brachytherapy taking the three-dimensional treatment geometry into account increases the dosimetric accuracy in planning and follow up of treatments. The comprehensive description and derivations provided gives a rigid background for further clinical, educational and research applications.

Results from the second Galaxy Serpent web-based table top exercise utilizing the concept of nuclear forensics libraries

Galaxy Serpent is a unique, virtual, web-based international tabletop series of exercises designed to mature the concept of National Nuclear Forensics Libraries (NNFLs). Teams participating in the second version of the exercise were provided synthetic sealed radioactive source data used to compile a model NNFL which then served as a comparative instrument in hypothetical scenarios involving sources out of regulatory control, allowing teams to successfully down-select and determine whether investigated sources were consistent with holdings in their model library. The methodologies utilized and aggregate results of the exercise will be presented, along with challenges encountered and benefits realized.

Study of Classification and Disposal Method for Disused Sealed Radioactive Source in Korea

Study of Classification and Disposal Method for Disused Sealed Radioactive Source in Korea

SIMPLIFIED PRACTICAL TEST METHOD FOR PORTABLE DOSE METERS USING SEVERAL SEALED RADIOACTIVE SOURCES

Sealed radioactive sources which have small activity were employed for the determination of response and tests for non-linearity and energy dependence of detector responses. Close source-to-detector geometry (at 0.3 m or less) was employed to practical tests for portable dose meters to accumulate statistically sufficient ionizing currents. Difference betweenresponse in the present experimentally studied field and in the reference field complied with ISO 4037 due to non-uniformity of radiation fluence at close geometry was corrected by use of Monte Carlo simulation. As a consequence, corrected results were consistent with the results obtained in the ISO 4037 reference field within their uncertainties.

光ルミネッセンス測定装置への密封 β 線源の導入と放射線管理

光ルミネッセンス測定装置への密封 β 線源の導入と放射線管理

Radiation protection in the design of γ-ray industrial computed tomography systems

The recent increasing use of γ-ray industrial computed tomography (γ-ray ICT) in various fields has induced greater attention to its performance as well as to considerations of radiation safety. It is understood that radiation protection planning cannot be sacrificed for the sake of CT image quality during the design, manufacture, and layout of γ-ray ICT systems. In the present work, we describe a typical γ-ray ICT system in brief, and, based on experience and pertinent examples, we propose design requirements for ensuring the radiation safety of the sealed radioactive source, source container, and workspace. The design examples and dose rate measurement results illustrate that the proposed design standards are reasonable, feasible, and safe, and are therefore meaningful for the design, manufacture, and layout of γ-ray ICT systems. This paper discussed the predominant measures associated with the radiation protection of γ-ray ICT systems in accordance with the pertinent Chinese standards. In addition, based on experience and pertinent examples, the design requirements for ensuring the radiation safety of a sealed radioactive source, source container, and workspace were defined in detail. The design examples and dose rate measurements conducted in conjunction with a γ-ray ICT system and workspace employing the proposed design standards have illustrated that the proposals provided in this paper are reasonable, feasible, and safe, and are therefore meaningful for the design, manufacture, and layout of γ-ray ICT systems.

Evaluation of a deterministic grid-based Boltzmann solver (GBBS) for voxel-level absorbed dose calculations in nuclear medicine

To evaluate the 3D Grid-based Boltzmann Solver (GBBS) code ATTILA® for coupled electron and photon transport in the nuclear medicine energy regime for electron (beta, Auger and internal conversion electrons) and photon (gamma, x-ray) sources. Codes rewritten based on ATTILA are used clinically for both high-energy photon teletherapy and 192Ir sealed source brachytherapy; little information exists for using the GBBS to calculate voxel-level absorbed doses in nuclear medicine.We compared DOSXYZnrc Monte Carlo (MC) with published voxel-S-values to establish MC as truth. GBBS was investigated for mono-energetic 1.0, 0.1, and 0.01 MeV electron and photon sources as well as 131I and 90Y radionuclides. We investigated convergence of GBBS by analyzing different meshes (), energy group structures () for each radionuclide component, angular quadrature orders (, and scattering order expansions (–); higher indices imply finer discretization. We compared GBBS to MC in (1) voxel-S-value geometry for soft tissue, lung, and bone, and (2) a source at the interface between combinations of lung, soft tissue, and bone.Excluding Auger and conversion electrons, MC agreed within ≈5% of published source voxel absorbed doses. For the finest discretization, most GBBS absorbed doses in the source voxel changed by less than 1% compared to the next finest discretization along each phase space variable indicating sufficient convergence. For the finest discretization, agreement with MC in the source voxel ranged from −3% to −20% with larger differences at lower energies (−3% for 1 MeV electron in lung to −20% for 0.01 MeV photon in bone); similar agreement was found for the interface geometries. Differences between GBBS and MC in the source voxel for 90Y and 131I were −6%.The GBBS ATTILA was benchmarked against MC in the nuclear medicine regime. GBBS can be a viable alternative to MC for voxel-level absorbed doses in nuclear medicine. However, reconciliation of the differences between GBBS and MC at lower energies requires further investigation of energy deposition cross-sections.

Optimization of energy window for gamma densitometer based backscatter method in oil industry

In this paper, we describe the results of various calculations performed for a design of the interference gauges that use the gamma-ray backscattering method with different scatter windows. The Pb barrier as shielding placed between 4 mCi 137Cs sealed radioactive source and a NaI (Tl) scintillation detector (1 inch in diameter—2 inch in height). Upon interaction with the material the scattered radiation, due to Compton process, was measured by NaI(Tl) detector. Using Compton scattered photons and review of the scattered photons in substance, by measuring of gamma-ray photons distributed, the density of materials can be obtained accurately. The results present for different scatter energy windows which 39–273 and 39–390 keV scatter windows have the best ΔX, STD and signal to noise.

Characterization of Radioisotope and Shield Content of Stored Disused Sealed Radioactive Sources in Ghana

AbstractRadioactive sources have been in use in Ghana over the last six decades, mainly in medicine, agriculture, industry, research, and academic investigations. Radioactive sources declared “disused” by users are required to be transferred to the Centralized Waste Processing and Storage facility of the National Radioactive Waste Management Center (NRWMC) located on the Ghana Atomic Energy Commission (GAEC) site in Kwabenya, a suburb of Accra. The NRWMC is the only national center mandated by regulation to receive and manage all radioactive waste in Ghana. A considerable number of radioactive sources received at the facility since its operation were without the necessary source information such as radionuclide, activity and reference date, identification number, supplier/manufacturer, etc. Sources received with information and those received without the proper information are both required to be properly characterized for the purpose of nuclear material accounting, segregation, pretreatment, conditioning...

Radioactivity determination of sealed pure beta-sources by surface dose measurements and Monte Carlo simulations

This study aims to determine the activity of a sealed pure beta-source by measuring the surface dose rate using an extrapolation chamber. A conversion factor (cGys−1Bq−1), which was defined as the ratio of surface dose rate to activity, can be calculated by Monte Carlo simulations of the extrapolation chamber measurement. To validate this hypothesis the certified activities of two standard pure beta-sources of Sr/Y-90 and Si/P-32 were compared with those determined by this method. In addition, a sealed test source of Sr/Y-90 was manufactured by the HANARO reactor group of KAERI (Korea Atomic Energy Research Institute) and used to further validate this method. The measured surface dose rates of the Sr/Y-90 and Si/P-32 standard sources were 4.615×10−5cGys−1 and 2.259×10−5cGys−1, respectively. The calculated conversion factors of the two sources were 1.213×10−8cGys−1Bq−1 and 1.071×10−8cGys−1Bq−1, respectively. Therefore, the activity of the standard Sr/Y-90 source was determined to be 3.995kBq, which was 2.0% less than the certified value (4.077kBq). For Si/P-32 the determined activity was 2.102kBq, which was 6.6% larger than the certified activity (1.971kBq). The activity of the Sr/Y-90 test source was determined to be 4.166kBq, while the apparent activity reported by KAERI was 5.803kBq. This large difference might be due to evaporation and diffusion of the source liquid during preparation and uncertainty in the amount of weighed aliquot of source liquid. The overall uncertainty involved in this method was determined to be 7.3%. We demonstrated that the activity of a sealed pure beta-source could be conveniently determined by complementary combination of measuring the surface dose rate and Monte Carlo simulations.

End of Life Decisions for Sealed Radioactive Sources

Sealed radioactive sources are encountered in a wide variety of settings-from household smoke detectors and instrument check sources through fixed industrial gauges, industrial radiography, and well logging sources, to irradiators and medical teletherapy devices. In general, the higher the level of activity in the sealed source, the stricter the regulatory control that is applied to its use, control, and ultimate disposition. Lower levels of attention and oversight can and do lead to sources ending up in the wrong place--as orphan sources in uncontrolled storage, disposed in a sanitary landfill, melted down in metal recycling operations and incorporated into consumer products, or handled by an unsuspecting member of the public. There is a range of issues that contribute to the problem of improper disposal of sealed sources and, in particular, to disused source disposal. Generally licensed sources and devices are particularly at risk of being disposed incorrectly. Higher activity generally licensed sources, although required to be registered with the (NRC) or an Agreement State, receive limited regulatory oversight and are not tracked on a national scale. Users frequently do not consider the full life-cycle costs when procuring sources or devices and discover that they cannot afford and/or are unwilling to pay the associated costs to package, transport and dispose of their sources properly. The NRC requirements for decommissioning funding plans and financial assurance are not adequate to cover sealed source transport and disposal costs fully. While there are regulatory limits for storage of disused sources, enforcement is limited, and there are only limited financial incentives in a small number of states for owners to dispose of the sources. In some cases, the lack of availability of approved Type B shipping casks presents an additional barrier to sealed source disposal. The report of the Disused Sources Working Group does an excellent job of framing these issues (www.disusedsources.org/wp-content/uploads/2014/12/DSWG-Report-March-2014.pdf). This article reviews both the issues and the report's recommendations, which are designed to improve sealed source control and encourage proper disposal of disused sources.

Enhanced Radioactive Material Source Security.

Requirements for additional security measures for sealed radioactive sources have evolved since they were first implemented after the terrorist events of 11 September 2001. This paper will describe the sequence of those measures, commencing with the early orders issued by the U.S. Nuclear Regulatory Commission to the May 2013 adoption of 10 CFR Part 37, Physical Protections of Category 1 and Category 2 Quantities of Radioactive Material. Part 37 requirements will be discussed in detail, as the 37 NRC Agreement States, which regulate approximately 88% of the radioactive material licensees, will be required to enact by 19 March 2016. In addition to the Part 37 requirements, the paper will also highlight some of the other ongoing efforts of the U.S. Department of Energy's National Nuclear Security Administration's Global Threat Reduction Initiative and the Conference of Radiation Control Program Directors.

NCRP Program Area Committee 2: Operational Radiation Safety.

Program Area Committee 2 of the National Council on Radiation Protection and Measurements provides guidance for radiation safety in occupational settings in a variety of industries and activities. The Committee completed three reports in recent years covering recommendations for the development and administration of radiation safety programs for smaller educational institutions, requirements for self-assessment programs that improve radiation safety and identify and correct deficiencies, and a comprehensive process for effective investigation of radiological incidents. Ongoing work includes a report on sealed radioactive source controls and oversight of a report on radioactive nanomaterials focusing on gaps within current radiation safety programs. Future efforts may deal with operational radiation safety programs in fields such as the safe use of handheld and portable x-ray fluorescence analyzers, occupational airborne radioactive contamination, unsealed radioactive sources, or industrial accelerators.

Recycling of a Disused 137Cs Sealed source

Recycling of disused sealed radioactive sources (DSRS) is one of the milestones of radioactive waste management strategy that prompts the minimisation of the radiological hazard to human health and the environment. In this respect this study spotlights recovering and purification of 137Cs as a solution (with a radioactivity of 191.5 ± 3.9 MBq) from an old disused source. The method was designed to fulfil the regulations and legislations of radiation protection giving exposure doses of the operating staff within the international permissible limits. Separation of the chemical impurities was performed by hydroxide precipitation based on pH values and solubility constants. X-ray fluorescence analysis illustrated a removal efficiency of 99% for iron and chromium, whereas the concentration of nickel and manganese decreased to 80 and 7 mg L-1, respectively, in the purified 137Cs solution. The radiochemical recovery of the purification process was determined using gamma-spectroscopy and found to be 94.4%. The purified 137Cs solution was then used to prepare a gel-source with 20 mL cylindrical geometry. Calibration was carried out at a secondary standard laboratory “National Radioactive Measurement Laboratory” (NRML) that is traceable to the primary standard “National Physics Laboratory” in the UK (NPL), giving an activity of 10.59 MBq with relative uncertainty of 0.4-0.5%. The preparation procedure could be considered as an original, facile, feasible and cost effective method that implies beneficial reuse of a DSRS in terms of volume minimization of radioactive wastes.

Personnel dose assessment due to the normal operations with the artificial radiation sources according to the data from the unified system of individual dose control (USIDC)

The aim of the study was personnel dose assessment due to the normal operations with the artificial radiation sources. The article is based on the data from the Unified System of Individual Dose Control and Voronezh Region’s radiation-hygienic passport. The data from No.1-DOZ “Information on personnel exposure doses under normal operation of technogenic ionizing radiation sources” and over a period of 2006-2010 years were analyzed. In 2006-2015, the number of organizations, which submitted form No.1-DOZ “Information on personnel exposure doses under normal operation of technogenic ionizing radiation sources”, increased from 175 to 229. In amount of the radiation facilities, Novovoronezh Nuclear Power Plant is the first. Novovoronezh NPP has 1512 sources, which amounts to 51,9% from all sources in Voronezh Region (2915). Health care facilities have 869 radiation sources or 29,8%. X-ray machines are the main part of these sources (844 health care facilities or 97,1% of all medical sources). Industrial sources occupy third place with 305 facilities or 10,5% of all considered sources. In 2015, according to the data from Voronezh Region’s radiation-hygienic passport, the number of “A” group personnel were 4237, the number of “B” group personnel were 2341. The average individual dose for personnel was over the range from 0.66 to 2.02 mSv. Collective dose was from 4.16 to 11.79 man-sieverts per year. The increase of number of the radiation sources has attended with the decrease of individual and collective doses. The most likely it is related to using the modern facilities. In 2015, the maximum value of the average individual dose of “A” group personnel was registered in Voronezh regional hospital (6.17 mSv y–1). There are medical facilities with unsealed and sealed sources in this hospital. In 2006-2015, the average individual doses of personnel of all radiation facilities that use radiation sources in Voronezh region were significantly below the dose limits established by the Radiation Safety Standards NRB-99/2009.

Metrological tests of a 200 L calibration source for HPGE detector systems for assay of radioactive waste drums

In this work we present test procedures, approval criteria and results from two metrological inspections of a certified large volume 152Eu source (drum about 200L) intended for calibration of HPGe gamma assay systems used for activity measurement of radioactive waste drums. The aim of the inspections was to prove the stability of the calibration source during its working life. The large volume source was designed and produced in 2007. It consists of 448 identical sealed radioactive sources (modules) apportioned in 32 transparent plastic tubes which were placed in a wooden matrix which filled the drum. During the inspections the modules were subjected to tests for verification of their certified characteristics. The results show a perfect compliance with the NIST basic guidelines for the properties of a radioactive certified reference material (CRM) and demonstrate the stability of the large volume CRM-drum after 7 years of operation.