Potential of modern technologies for improving internal exposure monitoring

Internal dosimetry is the science of assessing the amount and distribution of radionuclides in the body, and calculating resulting radiation doses to internal organs or tissues over specific time periods. Because the ionizing radiation energy deposited in a particular organ from radionuclides incorporated in the body cannot be measured directly, internal doses are estimated or inferred principally from in vivo or in vitro bioassay. As a matter of fact, in an effort to implement effective programmes in internal dosimetry, since internal dosimetry programmes exist, the internal dosimetry laboratories have always tried to develop new capabilities for these techniques or achieve the harmonisation in individual monitoring for occupational exposures. The primary goal of this paper is to categorise the principal trends made in recent developments in these fields regarding their potential and eligibility for the routine monitoring community and discuss the main aspects, which aims at a comprehensive assessment of these techniques. Secondly, starting from these data, their potential improvements are compared to the currently employed monitoring techniques used in routines.     

Sensitivity of a low energy Ge detector system for in vivo monitoring in the framework of ICRP 78 applications

In in vivo detection of internal contamination by actinides the minimum detectable activities (MDAs) correspond to significant doses, so the sensitivity of the detection system is the key to establishing adequate individual monitoring programmes for internal exposure to these radionuclides. The whole body counting (WBC) faculty at CIEMAT uses a low-energy Ge detector system with different available counting geometries to estimate the retention of actinides in the lungs and evaluate 125I in thyroid and 241Am in bone (skull and knee). A study of the factors and uncertainties involved in estimations of MDA is presented for lung and thyroid monitoring. The dependence of detection limits on counting efficiency in the measurement of low-energy emitters in the lungs has been carefully studied, carrying out a comparison among different biometric equations obtained by ultrasound techniques for estimations of chest wall thickness. Dosimetric implications of the estimated MDAs are taken into account in the framework of ICRP 78 application and considering Spanish regulations. The main interest in lung measurements is for the assessment of occupational exposure. This work confirms the low-energy Ge detector system to be an adequate in vivo technique for the routine monitoring of internal exposure to most insoluble uranium compounds (detection of 3% enriched uranium in lungs), and also to be useful in special monitoring programmes or in the case of incidents when the detection of 241Am is required.     

Study of nuclear medicine practices in Portugal from an internal dosimetry perspective

Nuclear medicine practices involve the handling of a wide range of pharmaceuticals labelled with different radionuclides, for diagnostic and therapeutic purposes. This work intends to evaluate the potential risks of internal contamination of nuclear medicine staff in several Portuguese nuclear medicine services and to conclude about the requirement of a routine internal monitoring. A methodology proposed by the International Atomic Energy Agency (IAEA), providing a set of criteria to determine the need, or not, for an internal monitoring programme, was applied. The evaluation of the risk of internal contaminations in a given set of working conditions is based on the type and amount of radionuclides being handled, as well as the safety conditions with which they are manipulated. The application of the IAEA criteria showed that 73.1% of all the workers included in this study should be integrated in a routine monitoring programme for internal contaminations; more specifically, 100% of workers performing radioimmunoassay techniques should be monitored. This study suggests that a routine monitoring programme for internal exposures should be implemented in Portugal for most nuclear medicine workers.     

Triage, monitoring and dose assessment for people exposed to ionising radiation following a malevolent act

The part played by individual monitoring within the context of the overall response to incidents involving the malevolent use of radiation or radioactive material is discussed. The main objectives of an individual monitoring programme are outlined, and types of malevolent use scenario briefly described. Some major challenges facing those with responsibilities for planning the monitoring response to such an incident are identified and discussed. These include the need for rapid selection and prioritisation of people for individual monitoring by means of an effective triage system; the need for rapid initiation of individual monitoring; problems associated with monitoring large numbers of people; the particular difficulties associated with incidents involving pure-beta and alpha-emitting radionuclides; the need for techniques that can provide retrospective estimates of external radiation exposures rapidly and the need for rapid interpretation of contamination monitoring data. The paper concludes with a brief review of assistance networks and relevant international projects planned or currently underway.     

A methodology for auto-monitoring of internal contamination by 131I in nuclear medicine workers

The manipulation of 131I in Nuclear Medicine involves significant risks of internal contamination of the staff. In the event of an accidental contamination, or when the Radiological Protection Program includes routine individual monitoring of internal contamination, it is necessary to implement internal dose estimation through in vivo and in vitro bioassay techniques. Due to the huge extension of the Brazilian country, this type of monitoring becomes unfeasible if all measurements have to be performed at the institutes of the CNEN. Thus, if the Nuclear Medicine Centres (NMC) become able to conduct the monitoring of their employees, this skill would be of great significance. The methodology proposed in this work consists in a simple and inexpensive protocol for auto-monitoring the internal contamination by 131I, using the resources available at the NMC. In order to verify the influence of the phantom in the calibration factor for the measurement of 131I in thyroid, it was performed a comparison among a variety of phantoms commercially available, including the Neck-Thyroid Phantom developed in IRD. A protocol for performing in vivo and in vitro measurements by the NMC was established. The applicability of the individual monitoring techniques was also evaluated by comparing the detection limits with the derived limits associated with the annual dose limits for workers.     

Design and implementation of monitoring programmes for internal exposure (project OMINEX)

The costs of monitoring for internal exposure in the workplace are usually significantly greater than the equivalent costs for external exposure. Therefore, there is a need to ensure that resources are employed with maximum effectiveness. The EC-funded OMINEX (optimisation of monitoring for internal exposure) project is developing methods for optimising the design and implementation of internal exposure monitoring programmes. Current monitoring programmes are being critically reviewed, the major sources of uncertainty in assessed internal dose investigated, and guidance formulated on factors such as programme design, choice of method/techniques, monitoring intervals, and monitoring frequency. OMINEX will promote a common, harmonised approach to the design and implementation of internal dose monitoring programmes throughout the EU.     

Assessment of doses to the offspring of the Techa River cohort due to intakes of radionuclides by the mother

The Techa River was contaminated as a result of radioactive releases by the Mayak plutonium production facility in 1949-1956. The residents of riverside communities were exposed to internal irradiation from radionuclides ingested mainly with river water, and also to external gamma irradiation resulting from shoreline and flood-plain contamination. The most important role in population exposure was played by (89,90)Sr and 137Cs. The persons born after the onset of the contamination have been identified as the 'Techa River Offspring Cohort' (TROC). The TROC has the potential to provide direct data on health effects in progeny that resulted from exposure of a general population to chronic radiation. This report describes the results of the calculation of fetal doses due to intakes of radionuclides by their mothers. Particular attention has been given to fetal dose from 90Sr because this nuclide is the most significant in terms of population dose for the Techa River. The comparison of the fetal bone marrow doses evaluated using different approaches proposed in the literature has shown a large dispersal in dose values. The main cause of this is the difference in model assumptions simplifying some developmental aspects of fetal haematopoiesis and bone formation. This paper presents an analysis of these basic assumptions that could be useful for further improvements in fetal dosimetry.     

Internal dosimetry: enhancements in application, update on the IDEA project

Efforts in many internal dosimetry laboratories to increase the accuracy and speed of measurements, to improve detection limits and to reduce uncertainties have resulted in improved or new internal monitoring techniques, both in in vivo measurement and in bioassay analysis. The EC-funded IDEA project (internal dosimetry--enhancements in application) aims to investigate why most of these techniques have not yet entered routine monitoring programmes (a summary of these new techniques is given here, showing their potential improvements compared with the currently employed monitoring routines). The project further aims at a comprehensive assessment of these techniques and the enhancements necessary to bring them to broader acceptance with those performing routine monitoring.     

State-of-the-art dosimetric methods for internal and external exposures: conclusions of a EURADOS action

In radiation protection dosimetry the quantity of interest is the effective dose (E); the dose limit to an adult worker applies to the sum of the relevant doses from external exposures and the relevant committed effective doses from intakes of radionuclides, during the same period of time. A EURADOS study was carried out to investigate how the results from personal dosemeters for external radiation, from workplace monitoring and from monitoring of internal exposures can be combined into a consistent system of individual monitoring. The integration of dosimetric methods and data for external and internal radiation require the complete characterisation of the occupational exposure present at the workplace, and the availability of adequate equipment and tools for the assessment of effective dose. To evaluate the capability of services to accomplish this approach, a European Dosimetry Network has been established among 28 European countries through the respective contact-persons and their dosimetric facilities which collaborated with EURADOS providing relevant data about performance and legal aspects. The information collected was presented as a monograph in Radiation Protection Dosimetry in 2004. The more relevant conclusions of this study are presented here.     

Environmental contamination and external radiation dose rates from radionuclides released from the Fukushima nuclear power plant

To evaluate the environmental contamination and contributory external exposure after the accident at the Fukushima Nuclear Power Plant (FNPP), the concentrations of artificial radionuclides in soil samples from each area were analysed by gamma spectrometry. Six artificial radionuclides ((131)I, (134)Cs, (137)Cs, (129m)Te, (95)Nb and (136)Cs) were detected in soil samples around FNPP. Calculated external effective doses from artificial radionuclide contamination in soil samples around FNPP were 1.9-2.9 μSv h(-1) (8.7-17.8 mSv y(-1)) in Fukushima city on 22 March 2011. After several months, these calculated external effective doses were 0.25-0.88 μSv h(-1) (2.2-7.6 mSv y(-1)) in Fukushima city on 29 June 2011. The present study revealed that the detected artificial radionuclides around FNPP mainly shifted to long-lived radionuclides such as radioactive caesium ((134)Cs and (137)Cs) even though current levels are decreasing gradually due to the decay of short-lived radionuclides such as (131)I, (129m)Te, (95)Nb and (136)Cs. Thus, radiation exposure potency still exists even though the national efforts are ongoing for reducing the annual exposure dose closer to 1 mSv, the public dose limit. Long-term environmental monitoring around FNPP contributes to radiation safety, with a reduction in unnecessary exposure to the residents.     

Developments in internal monitoring techniques

In an effort to increase accuracy and speed, improve detection limits and reduce uncertainties in internal dosimetry, laboratories have developed improved or new internal monitoring techniques in both in vivo measurements and bioassay analyses. Most of these techniques have not yet entered routine monitoring programmes. This paper intends to summarise these new techniques, show their potential improvements compared to the currently employed monitoring routines and discuss the main aspects of the EC-funded IDEA project, which aims at a comprehensive assessment of these techniques and the enhancements necessary to bring them to broader acceptance in the routine monitoring community.     

Control of the risk of exposure to alpha emitting radionuclides in French nuclear power plants: example of Cattenom

Control of the risk of internal exposure of EDF PWR plant maintenance workers by alpha-emitting radioactive elements is based on identification and quantification of the contamination of the systems. In 2001, an experiment carried out at Cattenom Power Plant during a unit outage in the presence of a leaking fuel cladding, based on measurement of alpha-emitting radioactive elements, made it possible to determine a realistic particle resuspension coefficient. A resuspension coefficient of 10(-6) m(-1) was adopted for operational radiological protection. An appropriate monitoring system for workers was set in place in collaboration with the occupational medicine and radiological protection department. It was based on prior estimation of the level of alpha contamination, and confirmed by swipe measurements, atmospheric surveillance by monitors, and collective analysis by nose blow samples from workers selected on the basis of their workstations, as well as supplementary individual measurements (monitoring of faeces). This surveillance made it possible to validate an appropriate work area monitoring system.     

Electronic look-up tables on retention and excretion of radionuclides as a PC based support system for internal dosimetry

For the implementation of internal dosimetry in workplaces based on individual monitoring, information that relates the measurement results with the intake of radionuclides is required. The present work provides electronic look-up tables as a PC based support system for internal dosimetry. The tables contain the computed values for retention and excretion of 42 selected radionuclides following a single intake by inhalation and by ingestion, where the new ICRP respiratory tract model and the latest ICRP biokinetic models were used. These look-up tables contain the day-by-day data up to 1000 days and the data at every 10 days up to 10,000 days for the monitoring quantities. Users can readily evaluate intake of radionuclides by dividing the measured results with the values in these electronic look-up tables.     

Evaluation of internal contamination levels after a radiological dispersal device incident using portal monitors

Following a radioactive dispersal device (RDD) incident, it may be necessary to evaluate the internal contamination levels of a large number of potentially affected individuals to determine if immediate medical follow-up is necessary. Since the current laboratory capacity to screen for internal contamination is limited, rapid field screening methods can be useful in prioritising individuals. This study evaluated the suitability of a radiation portal monitor for such screening. A model of the portal monitor was created for use with models of six anthropomorphic phantoms in Monte Carlo N-Particle Transport Code Version 5 (MCNP) X-5 Monte Carlo Team (MCNP-A General Monte Carlo N-Particle Transport Code Version 5. LA-CP-03-0245. Vol. 2. Los Alamos National Laboratory, 2004.). The count rates of the portal monitor were simulated for inhalation and ingestion of likely radionuclides from an RDD for each of the phantoms. The time-dependant organ concentrations of the radionuclides were determined using Dose and Risk Calculation Software Eckerman, Leggett, Cristy, Nelson, Ryman, Sjoreen and Ward (Dose and Risk Calculation Software Ver. 8.4. ORNL/TM-2001/190. Oak Ridge National Laboratory, 2006.). Portal monitor count rates corresponding to a committed effective dose E(50) of 10 mSv are reported.     

The medical management of unintentional radionuclide intakes

As a general medical problem, radionuclide intakes that may cause significant health effects are uncommon events. In preparing to manage a radionuclide accident, planning is the key. The medical aspects of such an accident are only one part of the management, and a professional team approach is required. Specific priorities and sequencing are necessary in medically managing a radionuclide intake. As soon as is reasonably practical, promptly remove the victim(s) from further radionuclide, radiation field, or chemical exposure. Life and limb-saving medical aid takes precedence over ionising radiation concerns in nearly all cases. Next are the prevention and/or minimisation of internal intake of radionuclides and evaluation and control of external radionuclide contamination, followed by institution of treatment to minimise the retained radionuclide. Communication with the accident victim, and his or her family, and public affairs/media issues are important. Finally, follow-up treatment for internal intakes that may cause delayed health effects is given.     

Internal dosimetry: towards harmonisation and coordination of research

The CONRAD Project is a Coordinated Network for Radiation Dosimetry funded by the European Commission 6th Framework Programme. The activities developed within CONRAD Work Package 5 ('Coordination of Research on Internal Dosimetry') have contributed to improve the harmonisation and reliability in the assessment of internal doses. The tasks carried out included a study of uncertainties and the refinement of the IDEAS Guidelines associated with the evaluation of doses after intakes of radionuclides. The implementation and quality assurance of new biokinetic models for dose assessment and the first attempt to develop a generic dosimetric model for DTPA therapy are important WP5 achievements. Applications of voxel phantoms and Monte Carlo simulations for the assessment of intakes from in vivo measurements were also considered. A Nuclear Emergency Monitoring Network (EUREMON) has been established for the interpretation of monitoring data after accidental or deliberate releases of radionuclides. Finally, WP5 group has worked on the update of the existing IDEAS bibliographic, internal contamination and case evaluation databases. A summary of CONRAD WP5 objectives and results is presented here.     

Harmonisation (legal, dosimetric, quality aspects) of individual monitoring, and integration of monitoring for external and internal exposures (EURADOS Working Group)

The EURADOS Working Group II on 'Harmonisation of individual monitoring' consists of experts from almost all EU Member States and Newly Associated States (NAS), involved in tasks related to the assessment of doses for internal and external radiation. The final objective is to achieve harmonisation in individual monitoring for occupational exposures. Sub-group 2 activities are focused on investigating how the results from personal dosemeters for external radiation and workplace monitoring and from monitoring for internal exposure can be combined into a complete and consistent system of individual monitoring. Three questionnaires were prepared, covering 'Individual monitoring of external radiation' (Questionnaire 1), 'Internal exposure' (Questionnaire 2) and 'Natural sources of radiation at the workplace' (Questionnaire 3). With the agreement of a 'contact-person', selected in each country, the distribution of the three EURADOS 2002 questionnaires was carried out by e-mail among the dosimetry facilities of 28 European countries. The preliminary results of these actions are presented here.     

The ICRP protection quantities, equivalent and effective dose: their basis and application

Equivalent and effective dose are protection quantities defined by the The International Commission on Radiological Protection (ICRP). They are frequently referred to simply as dose and may be misused. They provide a method for the summation of doses received from external sources and from intakes of radionuclides for comparison with dose limits and constraints, set to limit the risk of cancer and hereditary effects. For the assessment of internal doses, ICRP provides dose coefficients (Sv Bq(-1)) for the ingestion or inhalation of radionuclides by workers and members of the public, including children. Dose coefficients have also been calculated for in utero exposures following maternal intakes and for the transfer of radionuclides in breast milk. In each case, values are given of committed equivalent doses to organs and tissues and committed effective dose. Their calculation involves the use of defined biokinetic and dosimetric models, including the use of reference phantoms representing the human body. Radiation weighting factors are used as a simple representation of the different effectiveness of different radiations in causing stochastic effects at low doses. A single set of tissue weighting factors is used to take account of the contribution of individual organs and tissues to overall detriment from cancer and hereditary effects, despite age- and gender-related differences in estimates of risk and contributions to risk. The results are quantities that are not individual specific but are reference values for protection purposes, relating to doses to phantoms. The ICRP protection quantities are not intended for detailed assessments of dose and risk to individuals. They should not be used in epidemiological analyses or the assessment of the possibility of occurrence and severity of tissue reactions (deterministic effects) at higher doses. Dose coefficients are published as reference values and as such have no associated uncertainty. Assessments of uncertainties may be appropriate in specific analyses of doses and risks and in epidemiological studies.