Use In Radiation Protection Research Articles

Utilizing bypass cement dust in the production of radiation shielding bismuth borate glass

Using the melting-quench technique, a new method for removing bypass cement dust (BCD) was developed in this study using an active environmental glassy system with the chemical formula (65–x)B2O3–5Li2O–5CdO–25BCD–xBi2O3, where x varied from 0 to 30 mol%. A glass density of 3.137 g/cm³ is attained without bismuth, while a density of 5.345 g/cm³ is obtained with 30 mol% Bi2O3. The unit peaks BO3 and BO4 in the FTIR spectra signify the presence of the BLCDBi-series structure. Elevated Bi2O3 concentrations augment BO4 units. Defects influence the optical absorbance of glass within the wavelength range of 190–1000 nm. These defects in the glass network are created by silicon electrons, silicon hole centers, and BCD as well as Bi3+ or Bi6+ after the addition of bismuth oxide to the glassy system. Eg values were found to decrease as Bi2O3 concentration increased, whereas EU and n values exhibited the opposite behavior. A gamma spectroscopy system and the Phy-X/PSD software are used to calculate the linear attenuation coefficient (LAC), mass attenuation coefficient (MAC), half-value layer (HVL), tenth-value layer (TVL), effective atomic number (Zeff), and effective electron density (Neff) of samples. In addition, the Phy-X/PSD software demonstrated good agreement between experimental and theoretical interpretations. Overall, the results showed that the BLCDBi-30 sample is suitable for use in radiation protection.

A good balance between the efficiency of ionizing radiation shielding and mechanical performance of various tin-based alloys: Comparative analysis

The Sn-Bix (x = 30, 40, 50, 60 and 70%) radiation shielding alloys were prepared using a melting process of the alloying material in a ceramic crucible at 200 °C for 120 min. The structural characterization for synthesized samples was carried out using X-ray Diffraction technique (XRD). The hardness and elastic properties of the prepared alloys was measured using micro-indentation creep technology and tensile test machine respectively. XRD pattern indicated that that the crystal size of Sn phase is lowered by increasing Bi concentration in the Sn matrix. The hardness decreases as the dwell time increases. The hardness values and elastic properties (in terms of young modulus) were significantly improved due to the decrease in the crystallite size by increasing Bi content. The stress exponent (n) value increase by increasing Bi content which mean that the mechanical properties improved due to increment of resistance. Furthermore, the produced alloys' nuclear shielding ability was investigated. The μm values were calculated using MCNP simulation code and XCOM software over a broad energy range of 0.015 MeV–15 MeV with good agreement between them. Several characteristics are estimated in order to properly understand the researched alloy's radiation and properties of neutron shielding. The findings showed that a higher Bi concentration causes the crystal size to decrease, improving the radiation shielding and mechanical qualities. The alloy Sn-70% Bi has a maximum increase of vicker hardness, tensile strength, and young's modulus. The findings of the shielding parameters indicate that the alloys under study are efficient gamma shielding materials in contrast to other typical shielding materials and materials that have been examined recently. The highest mechanical efficiency and rather strong gamma-ray shielding capabilities are found in the Sn–70Bi alloy. Owing to its ability to effectively balance mechanical performance and shielding, the Sn–70Bi are approved for use in radiation protection. In addition, when compared to all other manufactured alloys, typical neutron shielding materials, and recently investigated materials, the results further demonstrate that the Sn–70Bi alloy has the best neutron attenuation capability. Furthermore, in terms of mass stopping power (MSP) and projected range (PR), the Sn–70Bi alloy demonstrates the most effective attenuation for alpha particles (He+2) and protons (H+1). These results imply that the Sn–70Bi alloy provide superior mechanical performance and nuclear shielding for a range of applications including industrial, medicinal, and nuclear waste storage in the future.

Korean-specific iodine S values for use in internal dosimetry

The use of iodine S values derived using the International Commission Radiological Protection (ICRP) phantoms may introduce significant bias in internal dosimetry for Koreans due to anatomical variability. In the current study, we produced an extensive dataset of Korean S values for selected five iodine radioisotopes (I-125, I-129, I-131, I-133, and I-134) for use in radiation protection. To calculate S values, we implemented Monte Carlo simulations using the Mesh-type Reference Korean Phantoms (MRKPs), developed in a high-quality/fidelity mesh format. Noticeable differences were observed in S value comparisons between the Korean and ICRP reference phantoms with ratios (Korean/ICRP) widely ranging from 0.16 to 6.2. The majority of S value ratios were lower than the unity in Korean phantoms (interquartile range = 0.47-1.28; mean = 0.96; median = 0.69). The S values provided in the current study will be extensively utilized in iodine internal dosimetry for Koreans.

Synthesis of pristine CaZrO3 and CaZrO3/Pr6O11 ceramic samples and assessment of their radiation protection features

The past few decades have witnessed significant variations in the use of shielding materials and ionizing radiation. The use of ceramics as shields for protection against radiation has attracted interest in research owing to exceptional properties that distinguish them from other materials that are available to this end. Calcium zirconate (CaZrO3) is a lead-free ceramic material that exhibits good physicochemical properties, and is considered to be a good host for lanthanide elements, such as praseodymium ions, where this makes it a suitable alternative to lead and lead-based compounds as well as a promising candidate for use in radiation protection. In this study, we examine the radiation shielding properties of pristine CaZrO3 and Pr6O11/CaZrO3 ceramic samples by using Monte Carlo simulations. The formation of the desired phase was identified by the XRD technique. The data analysis indicated that the samples crystallized into an orthorhombic structure. The densities of the pristine CaZrO3 and the Pr6O11/CaZrO3 samples were 4.20 g/cm3 and 4.54 g/cm3, respectively. The analysis of their radiation shielding properties indicated an improvement in their linear attenuation coefficient (LAC), with values of 59.3%, 24.4%, 11.3%, and 8.1% at γ-photon energies of 0.121 MeV, 0.356 MeV, 0.662 MeV, and 2.506 MeV, respectively, when Pr6O11 was added to CaZrO3. The improvement in the LAC of the prepared samples was accompanied by a considerable reduction in their half-value thickness, lead-equivalent thickness, and transmission factor.

The scientific basis for the use of the linear no-threshold (LNT) model at low doses and dose rates in radiological protection

The linear no-threshold (LNT) model was introduced into the radiological protection system about 60 years ago, but this model and its use in radiation protection are still debated today. This article presents an overview of results on effects of exposure to low linear-energy-transfer radiation in radiobiology and epidemiology accumulated over the last decade and discusses their impact on the use of the LNT model in the assessment of radiation-related cancer risks at low doses. The knowledge acquired over the past 10 years, both in radiobiology and epidemiology, has reinforced scientific knowledge about cancer risks at low doses. In radiobiology, although certain mechanisms do not support linearity, the early stages of carcinogenesis comprised of mutational events, which are assumed to play a key role in carcinogenesis, show linear responses to doses from as low as 10 mGy. The impact of non-mutational mechanisms on the risk of radiation-related cancer at low doses is currently difficult to assess. In epidemiology, the results show excess cancer risks at dose levels of 100 mGy or less. While some recent results indicate non-linear dose relationships for some cancers, overall, the LNT model does not substantially overestimate the risks at low doses. Recent results, in radiobiology or in epidemiology, suggest that a dose threshold, if any, could not be greater than a few tens of mGy. The scientific knowledge currently available does not contradict the use of the LNT model for the assessment of radiation-related cancer risks within the radiological protection system, and no other dose-risk relationship seems more appropriate for radiological protection purposes.

Trends in radiation protection use and radiation exposure of European endourologists: an ESUT-YAU Endourology Group prospective trial

Trends in radiation protection use and radiation exposure of European endourologists: an ESUT-YAU Endourology Group prospective trial

An assessment of radiation use, knowledge and concerns in trainee cardiologists in Ireland.

To evaluate knowledge, routine use and concerns of trainee cardiologists in the Republic of Ireland regarding radiation use in the cardiac catheterization laboratory. We handed out a Radiation Questionnaire to cardiology trainees in February 2020 at the Irish Cardiac Society "Spring Meeting". The questionnaire assessed radiation protection use amongst trainees and tested knowledge of X-ray basics. Many trainees report inadequate access to properly sized lead protection, and infrequent dosimeter usage. Over one-third of trainees report musculoskeletal issues from wearing leads, the majority of whom use correct size lead <60% of the time.33.3% report radiation concerns will affect their decision making regarding subspeciality training, but notably 83.3% of females and only 19% of males surveyed report this, showing this is a bigger issue for females in cardiology. Less than half of trainees feel adequately educated about radiation. Our assessment show deficiencies in the provision and use of personal protective equipment to trainees, highlights extra radiation concerns of female cardiology trainees, and notes gaps in knowledge in radiation use. Our assessment highlights deficiencies in the education of cardiology trainees regarding ionizing radiation, and suggests this area needs to be improved upon.

Calibration of neutron dosimeters for radiation protection use at the ALBA synchrotron experimental hall

Bright beams of synchrotron radiation are produced from a 3 GeV electron accelerator at the Spanish synchrotron facility ALBA. The electron acceleration and storage mechanisms lead to complex secondary radiation fields. These fields include electrons, photons and neutrons of a wide energy range, and have an important pulsed behaviour and very significant spatial variations. In order to perform correct dose assessments for radiation protection purposes, it is necessary to measure separately electrons, photons and neutrons. In the case of neutron dosimetry, given the strong dependence with energy of the radiation weighting factor and of the fluence to ambient dose equivalent conversion coefficient, it is necessary to have some spectrometric information to correctly assess dose. Neutron dosimeters designed at UAB, based on PADC (Poly Allyl Diglycol Carbonate) track detectors, were exposed in different places of the ALBA experimental hall and tunnel. As to date, no calibration facility provides neutron fields similar to those found at ALBA, ambient dose equivalent was originally evaluated with calibration coefficients obtained from calibration at an Am-Be neutron source. A procedure for calibrating the dosimeters based on the knowledge, either by simulation or experimental determination, of the neutron unit fluence distribution ϕE(E) at every specific point of measurement and of the dosimeter response function in terms of fluence Rφ(E) is stablished in this work. Specific values of the calibration coefficient Si, that allows determining the ambient dose equivalent value from trak density measurements, were calculated at different points of interest in ALBA and are in the range 35.8–258 cm-2mSv-1.

Implications of recent epidemiologic studies for the linear nonthreshold model and radiation protection

The recently published NCRP Commentary No. 27 evaluated the new information from epidemiologic studies as to their degree of support for applying the linear nonthreshold (LNT) model of carcinogenic effects for radiation protection purposes (NCRP 2018 Implications of Recent Epidemiologic Studies for the Linear Nonthreshold Model and Radiation Protection, Commentary No. 27 (Bethesda, MD: National Council on Radiation Protection and Measurements)). The aim was to determine whether recent epidemiologic studies of low-LET radiation, particularly those at low doses and/or low dose rates (LD/LDR), broadly support the LNT model of carcinogenic risk or, on the contrary, demonstrate sufficient evidence that the LNT model is inappropriate for the purposes of radiation protection. An updated review was needed because a considerable number of reports of radiation epidemiologic studies based on new or updated data have been published since other major reviews were conducted by national and international scientific committees. The Commentary provides a critical review of the LD/LDR studies that are most directly applicable to current occupational, environmental and medical radiation exposure circumstances. This Memorandum summarises several of the more important LD/LDR studies that incorporate radiation dose responses for solid cancer and leukemia that were reviewed in Commentary No. 27. In addition, an overview is provided of radiation studies of breast and thyroid cancers, and cancer after childhood exposures. Non-cancers are briefly touched upon such as ischemic heart disease, cataracts, and heritable genetic effects. To assess the applicability and utility of the LNT model for radiation protection, the Commentary evaluated 29 epidemiologic studies or groups of studies, primarily of total solid cancer, in terms of strengths and weaknesses in their epidemiologic methods, dosimetry approaches, and statistical modelling, and the degree to which they supported a LNT model for continued use in radiation protection. Recommendations for how to make epidemiologic radiation studies more informative are outlined. The NCRP Committee recognises that the risks from LD/LDR exposures are small and uncertain. The Committee judged that the available epidemiologic data were broadly supportive of the LNT model and that at this time no alternative dose-response relationship appears more pragmatic or prudent for radiation protection purposes.

P102] Maximum dose to the thyroid organ assessment using APD UNFORS EDD-30 for patients submitted to mammography examinations

Purpose We characterized the active personal dosimeter EDD-30 based on a set of tests: dose and dose rate linearity, energy dependency, reproducibility and angular dependency to use in radiation protection and access the maximum dose in thyroid organ in patients submitted to mammography screening examinations. Methods Measurements were performed with two UNFORS dosimeters: reference XiO detector and EDD-30, set together on the surface of a Perspex phantom, using the X-ray beam of a mobile and mammography units. EDD-30 readings, Hp”(10) were obtained using the EDD-30 in the surface of NM thyroid phantom placed on top of the trunk of the anthropomorphic phantom Alderson RANDO and using CIRS BR12 slabs (50% glandularity) to simulate compressed breast thicknesses (t) in range 20–80 mm. Set up: cranium caudal (CC) view with AOP STANDARD mode; Mean air kerma at the entrance surface of the neck in the thyroid location, K i 0 = Hp ″ ( 10 ) · F Hp ( 10 ) / Hp ′ ( 10 ) · F Angle · F Energy · J mGy / mSv , F’s are correction factors and J converts Hp(10) to incident air kerma, IAK; Mean incident air kerma in thyroid organ, K i T , was obtained using K i T = K i 0 exp ( - μ PMMA · x ) , x = 0,8 cm, μ PMMA – linear attenuation coefficient of PMMA for Compton scatter beam energy. To obtain the maximum dose to the thyroid organ, H T , we considered the following approximations: medio-lateral-oblique view with a 10% increase in H T and MGD, relative to CC view, and K i T ⩾ H T . Results APD is linear in the dose rates interval 13 mSv/h - 2.0 Sv/h with relative response, r = 1.22 ± 6% (k = 2). Energy response and angle dependence are in accordance with manufacturer data, r ≅ 1.2 in the energy range 28–40 keV; r ≈ 0.8 for angles above 60 degrees. Maximum normalized absorbed dose in thyroid for a CC view (Ht/mAs) range: 0.006–0.024 μ Gy/mAs; In bilateral exam: 0.05 maximum Relative Organ Dose for the thyroid, ROD (Ht/MGD) range: 0.02–0.12 %. Conclusions Using a very different method, we reproduced the results obtained by Sechopoulos (Monte Carlo Simulation), and Hatziioannou (TLD readings). Thyroid organ dose for a mammography screening examination of an average breast (t = 55 mm) is very low (3,4 μ Gy) corresponding to approximately 30 min of exposure to natural radiation. No thyroid protection is recommended.

무연 방사선 차폐 시트에 대한 몬테카를로 전산모사

방사선 특히, 엑스선 또는 감마선으로부터 인체를 보호하기 위해 납(Pb)으로 된 보호 장구를 광범위하게 사용해왔다. 최근 납 중독 및 환경오염의 문제로 납을 대신하는 무연 방사선 차폐재의 개발이 활발히 이루어지고 있다. 차폐재의 성능 확보를 위해서는 제작 및 평가의 순환 사이클을 반복하게 된다. 본 연구는 실제 무연 방사선 차폐소재의 제작에 앞서 차폐재의 성능을 몬테카를로 전산모사를 통해 확인함으로써 가능한 차폐소재의 조합을 연구하였다. 방사선 차폐소재의 평가에 사용되는 조건으로 엑스선관을 Geant4를 이용하여 전산모사하고 획득된 광자 스펙트럼을 이용하여 텅스텐과 비스무스의 조합에 따른 차폐소재의 성능을 평가하였다. 차폐소재의 공극에 따른 성능 저하도 평가하였다. 방사선 차폐 소재 개발 시 공극률을 줄이는 것이 중요한 인자라는 것을 알 수 있었다. Radiation protection equipment has widely used to protect human body from radiations, for example X-ray and gamma ray. The material of the radiation protection equipment is mainly lead (Pb) which has brought out lead poisoning and pollution when the equipment is fallen into disuse. This problem makes research and development find new Pb-free materials for use of radiation protection. Manufacturing and evaluation processes for developing those material were carried out repletely until obtaining the performance of protection rate. In this study, combination possibility of shielding material was studied using Geant4 monte carlo simulation. X-ray tube under the same condition in the real measurement of the protection rate was simulated, and X-ray tube spectrum was obtained. The X-ray tube spectrum was applied to study on the protection rate and lead equivalent. The porosity effect was simulated, and was one of key factors to determine protection rate or lead equivalent in radiation protection sheet of Pb-free.

A Biokinetic Model for Systemic Nickel.

The International Commission on Radiological Protection (ICRP) is updating its suite of reference biokinetic models for internally deposited radionuclides. This paper reviews data for nickel and proposes an updated biokinetic model for systemic (absorbed) nickel in adult humans for use in radiation protection. Compared with the ICRP's current model for nickel, the proposed model is based on a larger set of observations of the behavior of nickel in human subjects and laboratory animals and provides a more realistic description of the paths of movement of nickel in the body. For the two most important radioisotopes of nickel, Ni and Ni, the proposed model yields substantially lower dose estimates per unit of activity reaching blood than the current ICRP model.

Features of calculating the effective atomic number of complex media by attenuating radiation of various energies

The method for calculating the effective atomic number Z эф of complex environments consisting of elements with very different atomic numbers has been proposed. In order to account for the strong nonlinear dependence of the total atomic attenuation coefficient from the atomic number of the element, in contrast to the well-known process, the exponent “ n ” is introduced in the calculation of the total atomic cross section per electron atom multielement material (electronic section). The verification of the method was carried out by comparing the values Z эф of the total atomic attenuation coefficient obtained with effective atomic number and calculated by the rule of additivity of atomic ratios. The significant improvement was shown in the accuracy of determining the effective atomic number of complex substances. Effective atomic numbers of the two-component samples Al - X (X = 26 Fe, 29 Cu, 40 Zr, 41 Nb, 42 Mo, 46 Pd, 47 Ag, 48 Cd, 49 In, 50 Sn) for four photon energies 40.32, 59.54, 94.52, and 136.55 keV were calculated using the atomic attenuation coefficient taken from XCOM table. The effective atomic number value were also obtained for the sample of binary composition Al - Sn at various weight concentrations Sn (0.1-0.9) and energy of 59.54 keV, and Al-Sn (~ 0.08) at four of the above energies. It has been shown that a significant impact on the obtained values is provided not only by the difference between the atomic numbers ( Z max - Z min ) of elements of the substance’s composition, but also by the ratio of the concentrations of these elements. The impact of the above factors was significantly reduced with the increasing photon energy because it reduced the contribution of the photoelectric absorption in the general weakening. The knowledge of the effective atomic number of the medium is necessary to assess the quality of the materials in relation to their use in radiation protection, medical diagnostics, in separation of rocks with different material composition for the operational control of the useful component’s mass fraction and others. Key words: X-ray radiation, attenuation coefficient, effective atomic number, binary alloys, widely differing atomic numbers DOI: http://dx.doi.org/10.15826/analitika.2017.21.2.005 A.N. Eritenko, A.L.Tsvetyanskii and A.A.Polev Southern Federal University, 105, ul. B. Sadovaia, Rostov-na-Donu, 344004, Russian Federation

Radiation shielding in dentistry: an update.

The purpose of this article was to review the literature and provide guidelines on the use of radiation protection for patients in the dental setting. There are limited published data on the effects of low radiation doses such as those used in dental radiology. Most of the evidence is subject to bias, with risk models extrapolated from higher dose models such as studies of the Hiroshima bomb survivors. However, the lack of evidence does not denote the absence of risk, as there is no established 'safe' level of radiation exposure. All imaging utilizing ionizing radiation carries a risk for the patient. Hence the patient benefits of imaging must outweigh the potential risk. All diagnostic imaging should adhere to three basic principles, these being justification, optimization and application of dose limits. This article discusses dose reduction techniques and shielding of sensitive organs, particularly the thyroid, during procedures such as intraoral imaging, orthopantomograms and imaging of the pregnant patient.

Estimate of the radiation source term for 18 F production via thick H 2 18 O targets bombarded with 18 MeV protons

Abstract The positron-emitting radionuclide most important from the point of view of radiation protection is 18 F. This isotope is usually produced by bombarding 18 O-enriched water with protons. Currently there are few experimental data on the radiation source term generated during these reactions. In addition, presently there is no theoretical estimates of this source term, for use in radiation protection, validated by experimental data. Up till now this term is calculated by using nuclear interactions’ simulation codes, such as ALICE91. An estimate of the energy spectra for neutrons and photons, induced by 18 MeV protons on H 2 18 O target, have been calculated by using MCNPX code with cross sections from release 0 of ENDF/B VII library for all materials except 18 O, for which TENDL-2012 library was used. This estimate was validated against a recent experiment carried out at the Japan Atomic Energy Agency (JAEA). The calculated spectra have generally well reproduced experiments. The results show that the calculated radiation source term may be used to estimate the neutron activation of the accelerator components and the cyclotron building, to calculate the cyclotron shielding, and to carry out radiation protection evaluations in general, for the case of cyclotrons producing 18 F by means of the 18 O(p,n) 18 F nuclear reactions, for proton energies up to 18 MeV.

Classical approaches to microdosimetry, with example of use in radiation protection, medicine and mechanistic understanding.

Microdosimetry is the study and application of the microscopic features and stochastics of ionising radiations that cannot be described by the average macroscopic quantity absorbed dose. Microdosimetry is intimately related to radiation quality, but it also encompasses the inhomogeneities of interactions and energy depositions within a single radiation type. A variety of past and current approaches will be summarised and some examples given of implications to radiation protection and medicine, as well as to basic mechanistic studies of radiation effects.

Radiation dose around a PET scanner installation: Comparison of Monte Carlo simulations, analytical calculations and experimental results

PurposeMonte Carlo study of radiation transmission around areas surrounding a PET room. MethodsAn extended population of patients administered with 18F-FDG for PET-CT investigations was studied, collecting air kerma rate and gamma ray spectra measurements at a reference distance. An MC model of the diagnostic room was developed, including the scanner and walls with variable material and thickness. MC simulations were carried out with the widely used code GEANT4. ResultsThe model was validated by comparing simulated radiation dose values and gamma ray spectra produced by a volumetric source with experimental measurements; ambient doses in the surrounding areas were assessed for different combinations of wall materials and shielding and compared with analytical calculations, based on the AAPM Report 108.In the range 1.5–3.0 times of the product between the linear attenuation coefficient and thickness of an absorber (μ x), it was observed that the effectiveness of different combinations of shielding is roughly equivalent. An extensive tabulation of results is given in the text. ConclusionsThe validation tests performed showed a satisfactory agreement between the simulated and expected results. The simulated dose rates incident on, and transmitted by the walls in our model of PET scanner room, are generally in good agreement with analytical estimates performed using the AAPM Publication No. 108 method. This provides an independent confirmation of AAPM's approach. Even in this specific field of application, GEANT4 proved to be a relevant and accurate tool for dosimetry estimates, shielding evaluation and for general radiation protection use.

Radiation Safety in Image-guided Spinal Injections

Fluoroscopy and computed tomography (CT) guidance for interventional procedures can provide clear vision of the targets. However, ionizing radiation during fluoroscopy and CT-guided procedures can cause serious damage to both patients and medical professionals, including skin lesions, cataracts, cancer, and teratogenic effects on fetus and genetic mutations. In spite of the potential radiation damages, recent studies show that worldwide physician interests and knowledge on radiation safety and appropriate use of radiation protection are lacking. More education and training are needed to enhance the physician awareness of radiation safety. Techniques such as utilizing less fluoroscopy time, increasing the distance from the radiation source, using pulsed fluoroscopy, lower dose, and collimation have been proven to decrease radiation exposure. Appropriate utilization of radiation protection equipment including under-table shield, mobile shield, and apron can potentially minimize the professional radiation exposure close to 0 even when hundreds of fluoroscopy-guided procedures are performed monthly. Techniques such as performing proper preprocedure planning, performing CT fluoroscopy intermittently, modifying the longitudinal scan length, number of scans, and tube current-exposure time can all minimize the radiation dose in CT-guided procedures. Development of surgical techniques such as a computer-assisted navigation for pedicle screw placement and stereotactic guidance system under fluoroscopy may further decrease radiation exposure to both patients and physicians. More research on ultrasound-guided spine procedures may bring a new technique to deliver the medications safely and accurately to the patients without any risks of ionizing radiation.

Radiation Safety and Education in the Applicants of the Final Test for the Expert of Pain Medicine

BackgroundThe C-arm fluoroscope is known as the most important equipment in pain interventions. This study was conducted to investigate the completion rate of education on radiation safety, the knowledge of radiation exposure, the use of radiation protection, and so on.MethodsUnsigned questionnaires were collected from the 27 pain physicians who applied for the final test to become an expert in pain medicine in 2011. The survey was composed of 12 questions about the position of the hospital, the kind of hospital, the use of C-arm fluoroscopy, radiation safety education, knowledge of annual permissible radiation dose, use of radiation protection, and efforts to reduce radiation exposure.ResultsIn this study, although most respondents (93%) had used C-arm fluoroscopy, only 33% of the physicians completed radiation safety education. Even though nine (33%) had received education on radiation safety, none of the physicians knew the annual permissible radiation dose. In comparing the radiation safety education group and the no-education group, the rate of wearing radiation-protective glasses or goggles and the use of radiation badges or dosimeters were significantly higher in the education group. However, in the use of other protective equipment, knowledge of radiation safety, and efforts to reduce radiation exposure, there were no statistical differences between the two groups.ConclusionsThe respondents knew very little about radiation safety and had low interest in their radiation exposure. To make the use of fluoroscopy safer, additional education, as well as attention to and knowledge of practices of radiation safety are required for pain physicians.

A Physiological Systems Model for Iodine for Use in Radiation Protection

This paper summarizes the biokinetic database for iodine in the human body and proposes a biokinetic model for systemic iodine for use in dose assessments for internally deposited radioiodine. The model consolidates and extends existing physiological systems models describing three subsystems of the iodine cycle in the body: circulating inorganic iodide, thyroidal iodine (trapping and organic binding of iodide and synthesis, storage and secretion of thyroid hormones), and extrathyroidal organic iodine. Thyroidal uptake of inorganic iodide is described as a function of stable iodine intake (Y, µg day(-1)) and thyroidal secretion of hormonal iodine (S, µg day(-1)). Baseline parameter values are developed for reference adults with typical iodine intake. Compared with the current systemic biokinetic model of the International Commission on Radiological Protection (ICRP) for occupational intake of radioiodine, the proposed model predicts higher absorbed doses to the thyroid per unit uptake to blood for very short-lived iodine isotopes, similar absorbed doses to thyroid for iodine isotopes with half-life of at least a few hours, and substantially higher estimates of absorbed dose to stomach wall, salivary gland and kidneys for most iodine isotopes. Absorbed dose estimates for intravenous administration of radioiodine-labeled thyroid hormones based on the proposed model differ substantially in some cases from current ICRP values.