Effective Dose Coefficients for Intakes of Uranium Via Contaminated Wounds for Reference Adults.

ABSTRACT: Effective dose coefficients for intakes of uranium radionuclides via contaminated wounds have been calculated for reference adults following the procedures in the ICRP 103 series. The number of transformations in each source region for all members of the radioactive series from time of intake to 50 y post intake are calculated by coupling the NCRP 156 wound model to the ICRP 137 systemic models and ICRP 100 human alimentary tract model. Together with the ICRP 107 nuclear decay data for dosimetric calculations, the ICRP 133 specific absorbed fractions are implemented to calculate the radiation-weighted S coefficient deposited in each target organ or tissue from each transformation in each source region. Effective dose coefficients for different categories of intake materials via contaminated wounds are calculated for the three major uranium isotopes-238U, 235U, and 234U. Originating from the combined effects of the new absorbed fractions, biokinetic and dosimetric models, the new coefficients are generally reduced by a percentage of 23-28% as compared to the old ones. The new dose coefficients benefit the assessment of internal exposures for intakes of uranium via contaminated wounds in actual applications.

Whole organism concentration ratios in freshwater wildlife from an Australian tropical U mining environment and the derivation of a water radiological quality guideline value.

More than 10,000 whole organism concentration ratio (CRwo-water) values for freshwater wildlife were derived from radionuclide and stable element data representing an Australian tropical U mining environment. The CRwo-water values were summarised into five wildlife groups (bird, fish, mollusc, reptile and vascular plant). The summarised CRwo-water values represented 77 organism-element combinations. The CRwo-water values for U decay series elements were used in a tier 3 ERICA assessment. The assessment results were used to derive a water radiological quality guideline value (GV) for radiation protection of freshwater ecosystems in the context of the planned remediation of the Ranger U mine. The GV was an above-background water 226Ra activity concentration of 14 mBq L-1 (filtered fraction) or approximately 22 mBq L-1 (total fraction). The GV was based on the results of mollusc-bivalve as the limiting organism for the freshwater ecosystem.

Establishing Radiological Screening Levels for Defense-related Uranium Mine (DRUM) Sites on BLM Land Using a Recreational Future-use Scenario.

Thousands of former uranium mining sites in the United States, primarily in the southwestern states of Colorado, Arizona, New Mexico, Arizona, and Utah, are being identified and evaluated to assess their potential for causing public and environmental impacts. The common radiological contaminant of concern that characterizes these sites is naturally occurring uranium ore and associated wastes that may have been left behind postmining. The majority of these sites were abandoned and in general, are referred to as abandoned uranium mines, regardless of the government authority currently managing the land or in some cases, assigned responsibility for the oversight of assessment and remediation. The U.S. Department of Energy has identified over 4,000 defense-related uranium mine sites from which uranium ore was purchased by the U.S. government for nuclear defense programs prior to 1970. U.S. Department of Energy has established a program to inventory and perform environmental screening on defense-related uranium mine sites. The focus of this paper is the approximately 2,400 defense-related uranium mine sites located on federal land managed by the Bureau of Land Management and the U.S. Forest Service. This paper presents the results of an analysis to develop radiological screening criteria for U.S. Department of Energy's defense-related uranium mine sites that can be used as input to the overall ranking of these sites for prioritization of additional assessment, reclamation, or remedial actions. For these sites managed by Bureau of Land Management, public access is typically limited to short-term use, primarily for recreational purposes. This is a broad category that can cover a range of possible activities, including camping, hiking, hunting, biking, all-terrain vehicle use, and horseback riding. The radiological screening levels were developed by calculating the radiological dose to future recreational users of defense-related uranium mine sites assuming a future camper spends two weeks per year at the site engaged in recreational activities. Although a number of possible exposure pathways were included in this analysis (inhalation and ingestion of dust and soil, radon and progeny inhalation, and gamma radiation exposure from the soil), it is desirable as a practical matter to determine what gamma exposure rate would ensure that the annual acceptable exposure as determined by the regulatory authority will not be exceeded in the future. Because these sites are generally remote and located in semiarid environments, traditional exposure scenarios often applied in these types of analyses (e.g., subsistent farmers and ranchers), including exposure pathways for the ingestion of locally grown food products and water, were not considered relevant to short-term recreational use.

RADIATION DOSE MEASUREMENTS FOR HIGH-INTENSITY LASER INTERACTIONS WITH SOLID TARGETS AT SLAC.

A systematic study of photon and neutron radiation doses generated in high-intensity laser-solid interactions is underway at SLAC National Accelerator Laboratory. These laser-solid experiments are being performed using a 25 TW (up to 1 J in 40 fs) femtosecond pulsed Ti:sapphire laser at the Linac Coherent Light Source's (LCLS) Matter in Extreme Conditions (MEC) facility. Radiation measurements were performed with passive and active detectors deployed at various locations inside and outside the target chamber. Results from radiation dose measurements for laser-solid experiments at SLAC MEC in 2014 with peak intensity between 1018 and 7.1 × 1019 W cm-2 are presented.

Dose specification for ¹9²Ir high dose rate brachytherapy in terms of dose-to-water-in-medium and dose-to-medium-in-medium.

Dose calculation in high dose rate brachytherapy with (192)Ir is usually based on the TG-43U1 protocol where all media are considered to be water. Several dose calculation algorithms have been developed that are capable of handling heterogeneities with two possibilities to report dose: dose-to-medium-in-medium (Dm,m) and dose-to-water-in-medium (Dw,m). The relation between Dm,m and Dw,m for (192)Ir is the main goal of this study, in particular the dependence of Dw,m on the dose calculation approach using either large cavity theory (LCT) or small cavity theory (SCT). A head and neck case was selected due to the presence of media with a large range of atomic numbers relevant to tissues and mass densities such as air, soft tissues and bone interfaces. This case was simulated using a Monte Carlo (MC) code to score: Dm,m, Dw,m (LCT), mean photon energy and photon fluence. Dw,m (SCT) was derived from MC simulations using the ratio between the unrestricted collisional stopping power of the actual medium and water. Differences between Dm,m and Dw,m (SCT or LCT) can be negligible (<1%) for some tissues e.g. muscle and significant for other tissues with differences of up to 14% for bone. Using SCT or LCT approaches leads to differences between Dw,m (SCT) and Dw,m (LCT) up to 29% for bone and 36% for teeth. The mean photon energy distribution ranges from 222 keV up to 356 keV. However, results obtained using mean photon energies are not equivalent to the ones obtained using the full, local photon spectrum. This work concludes that it is essential that brachytherapy studies clearly report the dose quantity. It further shows that while differences between Dm,m and Dw,m (SCT) mainly depend on tissue type, differences between Dm,m and Dw,m (LCT) are, in addition, significantly dependent on the local photon energy fluence spectrum which varies with distance to implanted sources.

Using tea as an artificial urine in a Canadian performance testing program for fission/activation products.

In recent years, the National Calibration Reference Centre for Bioassay and In Vivo Monitoring (NCRC) at the Radiation Protection Bureau (RPB), Health Canada, has been conducting investigations with black tea to develop a matrix that can be used to replace urine in each of the following performance testing programs (PTP): (1) tritium, (2) carbon-14, (3) the DUAL (i.e., 3H/14C), and (4) fission/activation products (F/AP). A 1% tea solution with thimerosal, which had worked successfully for tritium, carbon-14, and the DUAL, was selected and tested for the F/AP PTP because of its similarity to urine in color and UV-VIS spectra. However, application of this tea to samples of the F/AP program containing 133Ba, 137Cs, 57Co, and 60Co produced precipitates, which was an unexpected result. Further experiments showed that replacement of thimerosal with an alcohol at about 5% eliminated the precipitation problem. The alcohol can be ethanol, methanol, or isopropanol. In the experiments, the 1% tea, preserved with alcohol, remained clear and stable for at least 100 d. The duration of each PTP for the NCRC is limited to 90 d. Application of the CNSC S-106 regulatory standard to the tea produced acceptable accuracy and precision results. It was concluded that a suitable tea matrix for the F/AP program had been found.

Practical ways to reduce radiation dose for patients and staff during device implantations and electrophysiological procedures.

Despite the advent of non-fluoroscopic technology, fluoroscopy remains the cornerstone of imaging in most interventional electrophysiological procedures, from diagnostic studies over ablation interventions to device implantation. Moreover, many patients receive additional X-ray imaging, such as cardiac computed tomography and others. More and more complex procedures have the risk to increase the radiation exposure, both for the patients and the operators. The professional lifetime attributable excess cancer risk may be around 1 in 100 for the operators, the same as for a patient undergoing repetitive complex procedures. Moreover, recent reports have also hinted at an excess risk of brain tumours among interventional cardiologists. Apart from evaluating the need for and justifying the use of radiation to assist their procedures, physicians have to continuously explore ways to reduce the radiation exposure. After an introduction on how to quantify the radiation exposure and defining its current magnitude in electrophysiology compared with the other sources of radiation, this position paper wants to offer some very practical advice on how to reduce exposure to patients and staff. The text describes how customization of the X-ray system, workflow adaptations, and shielding measures can be implemented in the cath lab. The potential and the pitfalls of different non-fluoroscopic guiding technologies are discussed. Finally, we suggest further improvements that can be implemented by both the physicians and the industry in the future. We are confident that these suggestions are able to reduce patient and operator exposure by more than an order of magnitude, and therefore think that these recommendations are worth reading and implementing by any electrophysiological operator in the field.

Assessing the dose values received by patients during conventional radiography X-ray examinations and the technical condition of the equipment used for this purpose.

BACKGROUND: X-ray examination is associated with patient exposure to ionizing radiation. Dose values depend on the type of medical procedure used, the X-ray unit technical condition and exposure conditions selected. The aim of this study was to determine the dose value received by patients during certain conventional radiography X-ray examinations and to assess the technical condition of medical equipment used for this purpose. MATERIAL AND METHODS: The study covered the total number of 118 conventional diagnostic X-ray units located in the Masovian Voivodeship. The methodology used to assess the conventional diagnostic X-ray unit technical condition and the measurement of the radiation dose rate received by patients are based on test procedures developed by the Department of Radiation Protection and Radiobiology of the National Institute of Public Health - National Institute of Hygiene (Warszawa, Poland) accredited for compliance with PN-EN 17025 standard by the Polish Centre for Accreditation. RESULTS: It was found that 84.7% of X-ray units fully meet the criteria set out in the Polish legislation regarding the safe use of ionizing radiation in medicine, while 15.3% of the units do not meet some of them. The broadest dose value range was recorded for adult patients. Particularly, during lateral (LATl) lumbar spine radiography the recorded entrance surface dose (ESD) values ranged from 283.5 to 7827 µGy (mean: 2183.3 µGy). CONCLUSIONS: It is absolutely necessary to constantly monitor the technical condition of all X-ray units, because it affects population exposure to ionizing radiation. Furthermore, it is essential to raise radiographers' awareness of the effects that ionizing radiation exposure can have on the human body.

NUSIMEP-7: uranium isotope amount ratios in uranium particles.

The Institute for Reference Materials and Measurements (IRMM) has extensive experience in the development of isotopic reference materials and the organization of interlaboratory comparisons (ILC) for nuclear measurements in compliance with the respective international guidelines (ISO Guide 34:2009 and ISO/IEC 17043:2010). The IRMM Nuclear Signatures Interlaboratory Measurement Evaluation Program (NUSIMEP) is an external quality control program with the objective of providing materials for measurements of trace amounts of nuclear materials in environmental matrices. Measurements of the isotopic ratios of the elements uranium and plutonium in small amounts, typical of those found in environmental samples, are required for nuclear safeguards and security, for the control of environmental contamination and for the detection of nuclear proliferation. The measurement results of participants in NUSIMEP are evaluated according to international guidelines in comparison to independent external certified reference values with demonstrated metrological traceability and uncertainty. NUSIMEP-7 focused on measurements of uranium isotope amount ratios in uranium particles aiming to support European Safeguards Directorate General for Energy (DG ENER), the International Atomic Energy Agency's (IAEA) network of analytical laboratories for environmental sampling (NWAL) and laboratories in the field of particle analysis. Each participant was provided two certified test samples: one with single and one with double isotopic enrichment. These NUSIMEP test samples were prepared by controlled hydrolysis of certified uranium hexafluoride in a specially designed aerosol deposition chamber at IRMM. Laboratories participating in NUSIMEP-7 received the test samples of uranium particles on two graphite disks with undisclosed isotopic ratio values n((234)U)/n((238)U), n((235)U)/n((238)U) and n((236)U)/n((238)U). The uranium isotope ratios had to be measured using their routine analytical procedures. Measurement of the major ratio n((235)U)/n((238)U) was obligatory; measurement of the minor ratios n((234)U)/n((238)U) and n((236)U)/n((238)U) was optional. Of the twenty-four institutes that registered for NUSIMEP-7, 17 have reported their results achieved by different analytical methods. The results of NUSIMEP-7 confirm the capability of laboratories in measuring n((234)U)/n((238)U), n((235)U)/n((238)U) and n((236)U)/n((238)U) in uranium particles of the size below 1 µm diameter. Furthermore, they underpin the recent advances in instrumental techniques in the field of particle analysis. In addition, feedback from the measurement communities from nuclear safeguards, nuclear security and earth sciences was collected in view of identifying future needs for NUSIMEP interlaboratory comparisons.

Certified reference materials and reference methods for nuclear safeguards and security.

Confidence in comparability and reliability of measurement results in nuclear material and environmental sample analysis are established via certified reference materials (CRMs), reference measurements, and inter-laboratory comparisons (ILCs). Increased needs for quality control tools in proliferation resistance, environmental sample analysis, development of measurement capabilities over the years and progress in modern analytical techniques are the main reasons for the development of new reference materials and reference methods for nuclear safeguards and security. The Institute for Reference Materials and Measurements (IRMM) prepares and certifices large quantities of the so-called "large-sized dried" (LSD) spikes for accurate measurement of the uranium and plutonium content in dissolved nuclear fuel solutions by isotope dilution mass spectrometry (IDMS) and also develops particle reference materials applied for the detection of nuclear signatures in environmental samples. IRMM is currently replacing some of its exhausted stocks of CRMs with new ones whose specifications are up-to-date and tailored for the demands of modern analytical techniques. Some of the existing materials will be re-measured to improve the uncertainties associated with their certified values, and to enable laboratories to reduce their combined measurement uncertainty. Safeguards involve the quantitative verification by independent measurements so that no nuclear material is diverted from its intended peaceful use. Safeguards authorities pay particular attention to plutonium and the uranium isotope (235)U, indicating the so-called 'enrichment', in nuclear material and in environmental samples. In addition to the verification of the major ratios, n((235)U)/n((238)U) and n((240)Pu)/n((239)Pu), the minor ratios of the less abundant uranium and plutonium isotopes contain valuable information about the origin and the 'history' of material used for commercial or possibly clandestine purposes, and have therefore reached high level of attention for safeguards authorities. Furthermore, IRMM initiated and coordinated the development of a Modified Total Evaporation (MTE) technique for accurate abundance ratio measurements of the "minor" isotope-amount ratios of uranium and plutonium in nuclear material and, in combination with a multi-dynamic measurement technique and filament carburization, in environmental samples. Currently IRMM is engaged in a study on the development of plutonium reference materials for "age dating", i.e. determination of the time elapsed since the last separation of plutonium from its daughter nuclides. The decay of a radioactive parent isotope and the build-up of a corresponding amount of daughter nuclide serve as chronometer to calculate the age of a nuclear material. There are no such certified reference materials available yet.

Preparation and application of steeps of tea as new simulations of urine for the performance testing programme of 14C.

(14)C is one of the radionuclides for which the Canadian Nuclear Safety Commission has developed performance testing programmes (PTPs). During the PTP exercises, clients receive samples of natural urine containing spiked radionuclides, for testing. In these programmes, urine has disadvantages. These include (1) slow collection times from donors, (2) unpleasant smell and (3) potential to transmit diseases. To assist in solving some of these problems, the Canadian National Calibration Reference Centre for Bioassay and In Vivo Monitoring has conducted research with tea solutions, to find simpler, safer and more readily available alternatives to urine. This paper provides a new technique by which steeps of black tea have been successfully prepared for the (14)C PTP. The results of tea solutions compared well with those of urine. It was concluded that tea steeps, of which the spectroscopic and colour quenching properties have been adjusted, do provide appropriate urine simulations, suitable for use in PTPs.

Neutron response characterisation of an OSL neutron dosemeter.

The neutron response was characterised for Al(2)O(3):C material coated with (6)Li(2)CO(3). Al(2)O(3):C material, an optically stimulated luminescent (OSL) dosemeter, which is sensitive to both photons and beta radiation but is mostly insensitive to neutron radiation. The Al(2)O(3):C material was coated with (6)Li(2)CO(3) to create a material sensitive to neutrons, beta and photon radiation. In this paper, the neutron response of this new OSL material (OSLN) will be compared with various other neutron measurement technologies in laboratory and field environments.

Quality assurance of testing methods in incorporation monitoring at the officially recognised incorporation measurement office at Jülich, Germany.

The systematic quality assurance (QA) and control of testing methods in incorporation monitoring consists of continual measures for internal QA and additional measures such as external laboratory controls. This includes among other aspects accuracy, precision and descriptions of the methods as well as the representation and timely availability of analytic results of measurements and internal dose assessment. At the officially recognised incorporation measurement office at Jülich, QA is performed for direct measurements (whole-body counter), indirect measurements with radiochemical testing methods of excretion samples and internal dose assessment.

Evaluation of the performance of two LiF:Mg,Ti and LiF:Mg,Cu,P dosemeters for extremity monitoring.

In this paper, the results aimed at assessing the performance of two varieties of LiF detectors (LiF:Mg,Ti and LiF:Mg,Cu,P) in photon fields relatively to reproducibility, detection threshold and angular dependence as defined in the ISO 12794 standard are presented. The fading properties and the limit of detection were also investigated for both materials. The results suggest that both LiF varieties are well suited for extremity monitoring. However, better fading properties of LiF:Mg,Cu,P when compared with LiF:Mg,Ti, combined with previous results relatively to energy dependence suggests that LiF:Mg,Cu,P dosemeters are better suited for extremity monitoring.

A survey of the optical hazards associated with hospital light sources with reference to the Control of Artificial Optical Radiation at Work Regulations 2010.

Workplace exposure to coherent and incoherent optical radiation from artificial sources is regulated under the Artificial Optical Radiation Directive (AORD) 2006/25/EC, now implemented in the UK under the Control of Artificial Optical Radiation at Work Regulations (AOR) 2010. These regulations set out exposure limit values. Implementing the AOR (2010 Health and Safety Statutory Instrument No 1140 www.legislation.gov.uk/uksi/2010/1140/pdf/uksi_20101140_en.pdf) requirements in a hospital environment is a potentially complex problem because of the wide variety of sources used for illumination, diagnosis and therapy. A survey of sources of incoherent optical radiation in a large hospital is reported here. The survey covers examples of office lighting, operating theatre lighting, examination lamps, and sources for ultraviolet phototherapy and visible phototherapies, including photodynamic therapy and neonatal blue-light therapy. The results of the survey are used to inform consideration of the strategy that a hospital might reasonably adopt both to demonstrate compliance with the AOR (2010) and to direct implementation effort.

New ICRP recommendations.

This paper provides a review of the 2007 recommendations of the International Commission on Radiological Protection (ICRP). These new recommendations take account of the latest biological and physical information and consolidate the additional guidance provided by ICRP since 1990. The changes to the scientific data are not substantial. ICRP has retained its fundamental hypothesis for the induction of stochastic effects of linearity of dose and effect without threshold and a dose and dose-rate effectiveness factor (DDREF) of 2 to derive nominal risk coefficients for low doses and low dose rates. While the overall detriment from low radiation doses has remained unchanged, ICRP has made adjustments to the values of the radiation and tissue weighting factors. In particular, the tissue weighting factor for breast has increased while that for gonads has decreased. There are some presentational changes to the system of protection. While ICRP has maintained the three fundamental principles--justification, optimisation of protection, and dose limitation-it has attempted to develop a more holistic approach to radiological protection covering all exposure situations--planned, existing and emergency--and all radiation sources, whether of natural or artificial origin. This approach should ensure that attention is focused on those exposures that can reasonably be controlled. It has also strengthened the principle of optimisation of protection with a particular emphasis on the use of constraints for planned exposure situations and reference levels for existing and emergency exposure situations. Dose constraints and reference levels are categorised into three bands which should assist in rationalising the many values of dose restrictions given in earlier ICRP publications. There are no changes to the dose limits. ICRP also indicates its intentions with respect to the development of further guidance on the protection of the environment. The fact that these new recommendations are more a matter of consolidation of previous ICRP recommendations and guidance should provide confidence that the system of protection established by and large in its present form several decades ago has reached a certain level of maturity. As such, no major changes to radiological protection regulations based on the 1990 recommendations should be necessary.

Effective dose assessment for workers in caves in the Czech Republic: experiments with passive radon detectors.

The new RAMARN system for radon volume activity measurement was developed in 2003 and has been in use since then. RAMARN system consists of a plastic chamber that is conically cylindrical in shape and about 0.5 l in volume; a bare Kodak LR 115 is located on the bottom of this diffusion chamber. The size was chosen to avoid the influence of deposited decay products of radon. Kodak has a spectrometric character--the tracks are visualized only for alphas with energies between 1 and 3 MeV that touch the foil; thus the effective volume has a lens shape. The response therefore corresponds to diffused radon and half of (218)Po born by radon gas decay. The experiments described below were conducted as one part of a routine methodology control, focused on classifying worker irradiation from natural ionizing radiation sources in show caves and in caves used for speleotherapy.