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Technical recommendations for monitoring individuals for occupational intakes of radionuclides


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The TECHREC project, funded by the European Commission, will provide Technical Recommendations for Monitoring Individuals for Occupational Intakes of Radionuclides. It is expected that the document will be published by the European Commission as a report in its Radiation Protection Series during 2016. The project is coordinated by the European Radiation Dosimetry Group (EURADOS) and is being carried out by members of EURADOS Working Group 7 (Internal Dosimetry). This paper describes the aims and purpose of the Technical Recommendations, and explains how the project is organised.
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G. Etherington1,*, P. Be
´rard2, E. Blanchardon3, B. Breustedt4, C. M. Castellani5, C. Challeton-de Vathaire3,
A. Giussani6, D. Franck3, M. A. Lopez7, J. W. Marsh1and D. Nosske6
Public Health England (PHE), Centre for Radiation Chemical and Environmental Hazards, Didcot, UK
Commissariat a
´nergie Atomique et aux E
´nergies Alternatives (CEA), France
Institut de Radioprotection et de Su
´aire (IRSN), France
Karlsruhe Institute of Technology (KIT), Germany
Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile (ENEA), Italy
Bundesamt fu
¨r Strahlenschutz (BfS), Germany
Centro de Investigaciones Energe
´ticas, Medioambientales y Tecnolo
´gicas (CIEMAT), Spain
*Corresponding author:
The TECHREC project, funded by the European Commission, will provide Technical Recommendations for Monitoring
Individuals for Occupational Intakes of Radionuclides. It is expected that the document will be published by the European
Commission as a report in its Radiation Protection Series during 2016. The project is coordinated by the European Radiation
Dosimetry Group (EURADOS) and is being carried out by members of EURADOS Working Group 7 (Internal Dosimetry).
This paper describes the aims and purpose of the Technical Recommendations, and explains how the project is organised.
A number of international organisations provide stan-
dards, guidance, advice, and scientific and technical in-
formation on topics related to monitoring individuals
for occupational intakes of radionuclides. However, no
single document presents a complete account of the
subject. The International Commission on Radiological
Protection (ICRP) provides biokinetic and dosimetric
models and data intended for the assessment of internal
doses resulting from radionuclide intakes, in a series of
. While these models and data are es-
sential tools for assessing internal doses, ICRP does not
provide comprehensive guidance or recommendations
on internal contamination monitoring, neither does it
provide practical guidance on the methods for assessing
internal doses from individual monitoring data, except
in the case of simple situations where the worker is
exposed to a single radionuclide, and a single measure-
ment is made using a single monitoring method.
Working Group 13 of ISO Technical Committee
85, Sub-Committee 2 (TC85/SC2) has provided inter-
national standards on monitoring and internal dose
(11 – 13)
, which provide a highly standar-
dised approach that is appropriate for a significant
fraction of cases where monitoring and internal dose
assessment is required in the event of occupational
intakes of radionuclides. However, the approach
is not appropriate for complex cases where multiple
measurements have been made using different moni-
toring methods.
IAEA and ICRU have also provided guidance on
specific topics
(14 – 17)
, while EURADOS has issued
guidelines on the specific issue of a structured ap-
proach to internal dose assessment
The TECHREC project aims to meet the need for a
comprehensive, unified account of the subject by provid-
ing a report that will present Technical Recommendations
for Monitoring Individuals for Occupational Intakes of
Radionuclides. It is expected that the report will be pub-
lished by the European Commission in its Radiation
Protection Series during 2016. It will be complementary
to an already-published report that presents technical
recommendations for monitoring individuals occupa-
tionally exposed to external radiation
. The project is
coordinated by the European Radiation Dosimetry
Group (EURADOS) and is being carried out by
members of EURADOS Working Group 7 (Internal
Dosimetry). It commenced in May 2014 and will be
completed in May 2016.
The report will describe the principles of the
subject, and provide a comprehensive Best Practice
Guide. The Technical Recommendations will take
account of all recent developments and will be fully
up-to-date. They will also take account of future
developments, such as the expected updates to the EU
2013 Basic Safety Standards (EU-BSS) that will take
account of developments embodied in the ICRP
Occupational Intakes of Radionuclides (OIR) report
. It is expected that the OIR report series (and,
by implication the Publication 103 dosimetry system),
#Crown copyright 2015.
Radiation Protection Dosimetry (2016), Vol. 170, No.14, pp. 812 doi:10.1093/rpd/ncv395
Advance Access publication 13 October 2015
will effectively be adopted for internal dosimetry
by national authorities in EU countries on or before
February 2018. The Technical Recommendations have
therefore been developed so that they will be equally
applicable both before and after adoption of the OIR
report series by national authorities.
The Technical Recommendations are intended to be
primarily informative in nature, providing guidance
and recommendations on best practice. They will not in
themselves make prescriptive or normative statements
about practices that must be adopted. Nevertheless,
the Technical Recommendations will make clear where
authorities or organisations (e.g. the EU EURATOM
Directive, ISO or ICRP) have specified that certain
methods, practices or conventions are mandatory
according to their own regulations or schemes.
The target audience includes internal dosimetry
services operating within the EU, as well as compe-
tent national and international authorities. The
Technical Recommendations are also expected to be
of interest to site operators who are responsible for ra-
diation protection programmes, radiation protection
experts who provide advice to site operators, manu-
facturers, laboratories providing bioassay services,
and government bodies aiming to harmonise regula-
tions and guidance.
The project is organised according to the work
package (WP) structure shown in Figure 1. The figure
also shows the main interactions among the WPs. WP1
developed a methodology for drafting the Technical
Recommendations, including specification of the source
documents to be taken into account, and quality assur-
ance criteria.
The most important source documents are sum-
marised in Figure 2. Some of these source documents
are currently under revision, as indicated by the
shaded boxes in the figure.
The contents of such source documents will not be
reproduced in detail, since this would result in a
report of excessive (and unnecessary) length. Rather,
the information that can be found in these documents
will be summarised, and guidance will be given on
how to make best use of this information.
An important feature of the project is the strong
emphasis placed on national and international con-
sultations (WP2), both with colleagues working in in-
ternal dosimetry and with stakeholders.
For the first consultation, selected members of
comments during January and February 2015. For the
Figure 1. TECHREC work package structure.
second (external stakeholder) stage of consultation, na-
tional contacts (one per country) were identified in the 28
EU countries, 5 non-EU European countries (Albania,
Norway, Russia, Switzerland and Ukraine), 6 countries
in Asia and America (Argentina, Brazil, Canada,
China, Japan and United States) and 7 international
organisations (ICRP, IAEA, ISO, PROCORAD, the
EURATOM Article 31 Expert Group and ILO). The na-
tional contacts were asked to identify stakeholders, and a
total of 32 national contacts provided answers. These
national contacts were from 22 EU countries (Belgium,
Bulgaria, Croatia, Czech Republic, Denmark, Finland,
France, Germany, Greece, Hungary, Italy, Ireland,
Lithuania, Malta, Poland, Portugal, Romania, Slovakia,
Slovenia, Spain, Sweden and UK), 4 non-EU European
countries and 6 non-European countries.
A list of external stakeholders at national level was
then generated, which included Internal Dosimetry
Services, Radiation Protection Services, Regulatory
Bodies and National Radiation Protection Associations.
This final List of Stakeholders contains 227 persons from
32 countries, and 8 ‘reference organisations’ and expert
groups (which includes a High Level Group of Qualified
Experts on internal dosimetry and radon monitoring).
The external stakeholder consultation took place
during March and April 2015. A complete ‘First
Draft’ was then produced in May 2015, taking into
account the comments received, and sent to the
European Commission for comment.
Topics addressed in the Technical Recommendations
(WP3) are summarised in Table 1.
As well as the topics that would be expected in
such a document (i.e. General Principles, Monitoring
Techniques, Monitoring Programmes, Routine and
Special Dose Assessment, Quality Assurance, Require-
ments for Internal Dosimetry Services), a number of
specialist topics are also addressed. These include:
monitoring and dosimetry for wound cases, monitor-
ing and dose assessment when decorporation therapy
treatment has been applied, dosimetry for intakes of
radon and its progeny, monitoring for first responders
in the event of a major accident at a nuclear facility,
and dosimetry for the specific application of assess-
ment of risks to health (rather than for radiation pro-
tection purposes).
The Recommendations will be developed further
during 2015, and a final stage of external stakeholder
consultation will be initiated in November 2015.
Figure 2. Source documents used in the development of the recommendations.
During the later phases of the project, an advanced
draft will be provided to the Article 31 Group of
Experts for their review and comment (WP4).
The final version of the Technical Recommendations
in publishable form will be submitted to the European
Commission in May 2016.
The TECHREC project is funded by the European
Commission (Directorate-General for Energy, DG
ENER), under Service Contract No. ENER/2014/
1. International Commission on Radiological Protection
(ICRP). Limits for intake of radionuclides by workers.
ICRP Publication 30, Parts 1–4 and Supplements. Ann.
ICRP. Pergamon Press (1979–1988).
2. Age-dependent doses to members of the public from
intake of radionuclides. ICRP Publication 56, Part
1. Ann. ICRP 20(2). Pergamon Press (1989).
3. Human respiratory tract model for radiological protec-
tion. ICRP Publication 66. Ann. ICRP 24(1– 3).
Pergamon Press (1994).
4. Age-dependent doses to members of the public from
intake of radionuclides: Part 2, Ingestion dose coeffi-
cients. ICRP Publication 67. Ann. ICRP 23(3/4).
Pergamon Press (1993).
5. Dose coefficients for intake of radionuclides by workers.
ICRP Publication 68. Ann. ICRP 24(4). Pergamon
Press (1994).
6. Age-dependent doses to members of the public from intake
of radionuclides: Part 3, Ingestion dose coefficients.ICRP
Publication 69. Ann. ICRP 25(1). Pergamon Press (1995).
7. Age-dependent doses to members of the public from intake
of radionuclides: Part 4, Inhalation dose coefficients.
ICRP Publication 71. Ann. ICRP 25(3/4). Pergamon
Press (1995).
8. Age-dependent doses to members of the public from
intake of radionuclides: Part 5 Compilation of ingestion
and inhalation dose coefficients. ICRP Publication 72.
Ann. ICRP 26(1). Pergamon Press (1996).
9. Individual monitoring for internal exposure of workers –
Replacement of ICRP Publication 54. ICRP Publication
78. Ann. ICRP 27 (3/4). Pergamon Press (1997).
Table 1. Chapter headings and topics addressed in the technical recommendations.
Chapter Topics
A. Purpose, context, scope and overview Purpose, context and scope of the technical recommendations
Overview of internal dosimetry and monitoring
B. General principles Dosimetric and operational quantities
Biokinetic and dosimetric models
Methodologies for the assessment of intakes of radionuclides:
fundamental aspects including bioassay functions
Dose assessment: basic principles (including dose coefficients)
C. Monitoring techniques Methods of individual and workplace monitoring
D. Monitoring programmes Design of individual monitoring programmes
E. Routine and special dose assessment Interpretation of monitoring data
Dose assessment and interpretation: in practice—routine
Dose assessment and interpretation: in practice—special
Monitoring and dosimetry for wound cases (and cutaneous
Monitoring and dose assessment in the event of ‘decorporation
F. Uncertainties Accuracy requirements and uncertainty analysis
G. QA and criteria for approval and accreditation Dose recording and reporting
Quality assurance and quality control—monitoring
Quality assurance and quality control—dose assessment
Accreditation/certification according to ISO/IEC standards
and training
Participation in national and international intercomparisons
H. Requirements for internal dosimetry services Requirements for internal dosimetry services
Annex I. Reference biokinetic and dosimetric models
Annex II. Example dose calculations
Annex III. Radon dosimetry for workers
Annex IV. Monitoring and internal dosimetry for first
responders in a major accident at a nuclear facility
Annex V. Internal dosimetry for assessment of risks to
health (e.g. compensation cases)
10. Human alimentary tract model for radiological protection.
ICRP Publication 100. Ann. ICRP 36(1–2). Elsevier
11. Radiation protection. Monitoring of workers occupation-
ally exposed to a risk of internal contamination with
radioactive material. ISO 20553:2006. ISO (2006).
12. Radiation Protection. Dose assessment for the monitor-
ing of workers for internal radiation exposure. ISO
27048:2011. ISO (2011).
13. Radiation Protection. Performance criteria for radio-
bioassay. ISO 28218:2010. ISO (2011).
14. IAEA Safety Guide. Assessment of occupation exposure
due to intakes of radionuclides. IAEA Safety Standard
Series No. RS-G-1.2. IAEA (1999).
15. Direct Methods for Measuring Radionuclides in the
Human Body, Safety Series No. 114. IAEA (1996).
16. Indirect Methods for Assessing Intakes of Radionuclides
Causing Occupational Exposure Safety Reports Series
18. IAEA (2000).
17. International Commission of Radiation Units and
Measurements. Direct determination of the body
content of radionuclides. ICRU Report 69, J. ICRU 3
(1) (2003).
18. Castellani, C. M., Marsh, J. W., Hurtgen, C.,
Blanchardon, E., Be
´rard, P., Giussani, A. and Lopez,
M. A. IDEAS Guidelines (Version 2) for the Estimation
of Committed Doses from Incorporation Monitoring
Data. EURADOS Report 2013-01 ISBN 978-3-943701-
03-6 (2013).
19. Technical Recommendations for Monitoring Individuals
Occupationally Exposed to External Radiation. Radiation
Protection RP160. Directorate-General for Energy and
Transport, Directorate H—Nuclear Energy, Unit H.4—
Radiation Protection.http://
ener/files/documents/160.pdf (last accessed 28 April
20. Occupational Intakes of Radionuclides, Parts 1 –5. ICRP
(in press).
... However, new methods or deviations from standard procedures that show an increased risk of significant committed dose are best accompanied by a monitoring programme [4]. There are various reference works for the dosimetry and interpretation of monitoring data, e.g. the IDEAS guidelines from EURADOS [5], ICRP-78 [6], IAEA RS-37 [7], ISO standards [8,9] or the technical recommendations from the EU [10]. For one single procedure step, the potential committed effective dose D from the intake of radionuclides can be estimated from the handled activity (A), the assumed intake fraction (or intake factor) a and the dose conversion coefficient (effective dose coefficient) e that converts the activity of internal emitters into committed effective dose (depending on the intake path and physical characteristics of the radioactive material): ...
... showed that an intake factor of 10 −7 can be assumed for standard diagnostic procedures in nuclear medicine [13]. It has also been shown that the use of a cabinet reduces the intake factor by one order of magnitude, and the use of glove boxes can reduce the intake factor by two orders of magnitude (protection safety factor f ps , see [8] and [10]). The assumption of an intake factor of 10 −4 and an effective dose coefficient of 5.7 µSv/Bq [14] for 223 Ra implies that the single handling of 1.75 MBq (resulting in an assumed intake of 175 Bq) would already lead to a potential committed effective dose of around 1 mSv. ...
... The data are taken from the electronic Annex accompaying the ICRP Occupational Intake of Radionuclides publications series [18]. An AMAD of 5 µm represents the standard assumption for the establishment of monitoring programs [4,6,7,9,10]. ...
An intake monitoring program covering more than half a year of clinical administration of Radium-223-dichloride for the palliative treatment of castration-resistant prostate cancer was carried out in the nuclear medicine department of the university hospital Bonn. Radioactivity in a total of 87 samples of gloves, air filters, faecal bioassays and face masks was measured and evaluated to assess the need for radiation protection measures for the medical staff. The main aim was to quantify or obtain an upper limit for the intake factor. An intake factor of 10 ⁻⁸ was measured when the preparation of patient doses took place in part in a laminar flow cabinet, which indicates an intake factor of 10 ⁻⁷ in more commonplace practice without a cabinet. The intake factor is therefore at the same level as other standard applications of unsealed sources in nuclear medicine. Our findings confirmed that masks are not required under any circumstances. However, the investigation also revealed that contamination risks, especially during the preparation of doses in syringes, should not be neglected. © 2019 Society for Radiological Protection. Published on behalf of SRP by IOP Publishing Limited. All rights reserved.
... Three basic principles can be applied to perform the ALARA principle. These principles can be listed as the shortest exposure time, the longest distance from the source and the optimum shielding material thickness [3,4]. Among the mentioned principles of ALARA, the term of shielding is the suitable protective barrier between the radiation source and the staff. ...
... Fig. 2 demonstrates the narrow beam geometry of the GEANT4 simulation, consisting of a point gamma source impinging on a slab of the glass. The gamma photon energies were defined in the re 4 ...
This study aimed to investigate the shielding performance of SrO-LiF-B2O3 glasses glass system for nuclear security applications. The MCNPX code (version 2.6.0) and GEANT4 are used to determine the shielding parameters and the dependence with the composition of each glass, as well as the influence of Cr2O3 additive. A wide-range of nuclear radiation shielding investigation for gamma-ray, proton particles, fast neutrons have been studied for five different types of glasses. The calculated values for mass attenuation coefficients (μm) were utilized to determine other vital shielding properties against gamma-ray radiation. Furthermore, some of the investigated parameters have been determined by using SRIM code and special calculation methods such as G-P fitting parameters for EBF and EABF calculation. The results showed that C25 glass with the highest Cr2O3 additive had a satisfactory capacity in nuclear radiation shielding.
... The LDRLs suggested here are built on surveys based on actual patients, using the dose indicators on the CT display, CTDI vol , and DLP to ensure that results are comparable and applicable for different scanners, sites, and regions of the LDRL values are extremely less than those adopted in other countries, or suggested in other studies. In Japan, LDRL ranges from 30 to 55 mGy and from 480 to 850 for the CTDI vol and DLP values, respectively, and in the UK, it ranges from 23 to 52 mGy and from 315 to 750 of the CTDI vol , and DLP values, respectively (Shrimpton et al., 2005: Protection, 2007: Etherington et al., 2016: Granata et al., 2019: Kanda et al., 2021. This suggests an excellent optimization for reducing the dose received by patients during CT scans, especially in pediatric CT brain scans. ...
Full-text available
Background: To date in Saudi Arabia, a limited number of studies conducted to assess radiation doses received by pediatrics in computed tomography (CT) brain procedure. National diagnostic reference levels (NDRL) have been established for adults, but neither NDRL’s nor Local diagnostic reference levels (LDRL) have been established for pediatric patients. Objective: This study aimed to assess radiation doses experienced by pediatric patients in CT brain procedure, and derive LDRLs. Materials and methods: The values of three radiological indexes: volume CT dose index (CTDI vol ) and dose-length product (DLP) were assessed. Then effective dose (ED) were estimated, and LDRLs are suggested for CT procedures based on data retrieved from 353 pediatric patients aged between 0 and 15 years old. LDRLs were estimated based on age and weight. Results: Built on 75 percent of the median distribution of the CTDI vol and DLP values, weight assemblage LDRL values ranged from 12.29 to 28.72 mGy and from 173.32 to 565.38, respectively, whereas age assemblage LDRL values ranged from 11.76 to 25.07 mGy and from 147.04 to 479.23, respectively. Conclusion: This study derived the typical CTDI vol , DLP, and ED received by pediatric patient during CT brain procedure in Saudi Arabia. Then, LDRLs were proposed based on age and weight for pediatric patients aged between 0 to 15 years old.
... The term of radiation protection is set on ALARA principle namely As Low As Reasonably Achievable, which means that even the very low dose should be avoided if it is not necessary. ALARA can be applied through three basic parameters namely time, distance and shielding [3,4]. Among the those three basic rules, the term of shielding is defined as a reduction of scatter radiation by using high atomic number material [5]. ...
The aim of this investigation was to present the gamma-ray and neutron shielding properties of different type of Ga additive Pd/Mn binary alloys. Therefore, mass attenuation coefficient (μm) values of different types of Pd-Mn-Ga alloys were calculated using the MCNPX simulation code in the 0.02–15MeV energy region. Afterwards, the obtained μm values have been utilized for determination of HVL, TVL, MFP and Zeff gamma shielding parameters for evaluation of the gamma radiation shielding abilities of alloy samples. Moreover, effective removal cross-sections (ΣR) of investigated alloy samples have been calculared. The results showed that both μm and Zeff values are increased as the Ga concentration increases. Thus, the study is indicates that 25Ga alloy sample owns distinguished protection ability to attenuate the gamma-ray radiation. The results also showed that 0Ga alloy sample which containing %0 of Ga has the highest ΣR value as a most effective alloy sample among the other samples for fast neutron shielding. To observe the shielding paerformance of superior sample, we have compared our results with previous available investigations in literature. It can be concluded that Ga additive for this type of binary alloys can help to increase the shielding gamma-ray performance.
... ALARA rule, namely As Low As Reasonably Achievable meaning that very low dose should be avoided if it is not necessary. The term of ALARA can be utilized through three basic parameters namely time, distance and shielding [3,4]. From those three rules, the term of shielding can be defined as a reduction of scatter radiation by using certain materials with high atomic number [5]. ...
We present the synthesis and nuclear radiation shielding characterization of glasses based on germanium and bismuth oxide prepared with the melt-quenching technique. The glasses were produced using six different compositions as follows (in wt%): 40.0GeO2-60.0PbO, 38.0GeO2-62.0Bi2O3, 31.344GeO2-41.900Bi2O3-26.756PbO, 33.334GeO2-33.333TeO2-33.333PbO, 20.6GeO2-41.9TeO2-17.4Nb2O5-20.1BaO, 41.98Bi2O3-48.26PbO-8.10Ga2O3-1.66BaO. Monte Carlo method (MCNPX code, v-2.6.0) has been utilized for the determination of mass attenuation coefficients (μ/ρ) of the synthesised six different glasses. The acquired mass attenuation coefficients (μ/ρ) have been used to determine the vital parameters for gamma-ray shielding namely half value layer (HVL), mean free path (MFP), tenth value layer (TVL), effective atomic number (Zeff), exposure buildup factor (EBF), energy absorption buildup factor (EABF), transmission factors (TF), respectively. Simultaneously, effective removal cross section (∑R) values for fast neutrons and Proton mass stopping power & proton projected range have been also calculated. The results showed that among all the investigated glasses, 41.98Bi2O3-48.26PbO-8.10Ga2O3-1.66BaO glass sample has the extra capability to reduce nuclear radiation as a shielding material.
... In 2017 Working Group 7 of the European Radiation Dosimetry Group (EURADOS WG7) has organized an intercomparison action for internal dose assessment for occupational exposures. Main aim of the intercomparison action was to test the practical applicability of the "Technical Recommendations for Monitoring Individuals for Occupational Intakes of Radionuclides" ("TECHREC Recommendations") [1] Among the four proposed cases, Case 1 (an artificial case of accidental inhalation of 60 Co) was focussed on the impact of the changes introduced by the most recent ICRP recommendations [2], including the new report series on Occupational Intakes of Radionuclides (OIR) [3][4][5]. Additionally, the analysis of the results submitted for Case 4 (a complex real case of accidental inhalation of 241 Am) provided useful feedbacks on the revised biokinetic models of ICRP. ...
Monitoring internal exposure to short-lived radionuclides is challenging, due to the frequent measurements required. ISO Standard 16637 and the Swiss Personal Dosimetry Ordinance describe a screening measurement (triage monitoring) conducted in the workplace to identify workers suspected of internal exposure. Based on a previous study that examined the feasibility of using several commonly found radiation monitors in Israel in a triage monitoring program, we conducted a pilot study towards the implementation of triage monitoring in nuclear medicine facilities in Israel. The pilot study was conducted while considering the current Israeli regulations and local safety culture. We implemented the triage monitoring program in three nuclear medicine facilities in Israel, with a total of 55 monitored workers. The pilot study consisted of two stages: a short-term stage conducted in the largest manufacture of radiopharmaceuticals in Israel and a long-term stage in two nuclear medicine departments in Israel. During the first stage of the study, participants were asked to conduct a daily measurement at the end of the workday and send a urine sample to the national internal dosimetry laboratory. The second stage lasted 5 months in a major hospital and 18 months in a regional hospital. The workers were asked to perform the measurement at the end of the shift and send a urine sample if a defined threshold had been crossed. The mean participation rate in the long-term stage(>70%) indicates that implementation of the triage monitoring program could be successful in Israel. Based on the findings of the study, practical recommendations are listed: suitable monitoring devices, allocating a monitoring location, time of measurement, training of the workers, record keeping and coordination with a certified dosimetry laboratory. The pilot study recommendations were submitted to the Israel Institute for Occupational Safety and Hygiene at the Ministry of Labor, Social Affairs and Social Services.
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This work presents an overview of the applications of retrospective dosimetry techniques in case of incorporation of radionuclides. The fact that internal exposures are characterized by a spatially inhomogeneous irradiation of the body, which is potentially prolonged over large periods and variable over time, is particularly problematic for biological and electron paramagnetic resonance (EPR) dosimetry methods when compared with external exposures. The paper gives initially specific information about internal dosimetry methods, the most common cytogenetic techniques used in biological dosimetry and EPR dosimetry applied to tooth enamel. Based on real-case scenarios, dose estimates obtained from bioassay data as well as with biological and/or EPR dosimetry are compared and critically discussed. In most of the scenarios presented, concomitant external exposures were responsible for the greater portion of the received dose. As no assay is available which can discriminate between radiation of different types and different LETs on the basis of the type of damage induced, it is not possible to infer from these studies specific conclusions valid for incorporated radionuclides alone. The biological dosimetry assays and EPR techniques proved to be most applicable in cases when the radionuclides are almost homogeneously distributed in the body. No compelling evidence was obtained in other cases of extremely inhomogeneous distribution. Retrospective dosimetry needs to be optimized and further developed in order to be able to deal with real exposure cases, where a mixture of both external and internal exposures will be encountered most of the times.
This paper provides a summary of the Education and Training (E&T) activities that have been developed and organized by the European Radiation Dosimetry Group (EURADOS) in recent years and in the case of Training Courses over the last decade. These E&T actions include short duration Training Courses on well-established topics organized within the activity of EURADOS Working Groups (WGs), or one-day events integrated in the EURADOS Annual Meeting (workshops, winter schools, the intercomparison participants' sessions and the learning network, among others). Moreover, EURADOS has recently established a Young Scientist Grant and a Young Scientist Award. The Grant supports young scientists by encouraging them to perform research projects at other laboratories of the EURADOS network. The Award is given in recognition of excellent work developed within the WGs' work programme. Additionally, EURADOS supports the dissemination of knowledge in radiation dosimetry by promoting and endorsing conferences such as the individual monitoring (IM) series, the neutron and ion dosimetry symposia (NEUDOS) and contributions to E&T sessions at specific events.
In contrast to external dosimetry, which can be based on measurement of operational quantities directly related to the effective dose (mSv), the assessment of doses due to intakes of radionuclides relies on the measurement of activities (Bq) and subsequent modelling. The methodology and the models are published by the International Commission on Radiological Protection (ICRP) and are later implemented in national regulations. This paper describes the general concepts for assessment of internal occupational exposure and diagnostic nuclear medicine applications using biokinetic and dosimetric models. The current changes in the models which are presented in the OIR series (Occupational Intakes of Radionucldes) of ICRP are outlined and information on the uncertainties in internal dose assessment are presented.
Full-text available
Dose assessment after intakes of radionuclides requires application of biokinetic and dosimetric models and assumptions about factors influencing the final result. In 2006, a document giving guidance for such assessment was published, commonly referred to as the IDEAS Guidelines. Following its publication, a working group within the European networks CONRAD and EURADOS was established to improve and update the IDEAS Guidelines. This work resulted in Version 2 of the IDEAS Guidelines, which was published in 2013 in the form of a EURADOS report. The general structure of the original document was maintained; however, new procedures were included, e.g. the direct dose assessment method for 3H or special procedure for wound cases applying the NCRP wound model. In addition, information was updated and expanded, e.g. data on dietary excretion of U, Th, Ra and Po for urine and faeces or typical and achievable values for detection limits for different bioassay measurement techniques.
The present report describes biokinetic data and the dosimetric models used for calculating age-dependent dose coefficients for intakes of radionuclides by ingestion and inhalation following their release into the environment. Dose coefficients for intakes by inhalation are based on the current ICRP Lung Model, which does not include age-dependent parameters. Data are presented in this report for radioisotopes of the following elements: hydrogen, carbon, strontium, zirconium, niobium, ruthenium, iodine, caesium, cerium, plutonium, americium and neptunium.
Conference Paper
The Bioassay Performance Criteria standard provides requirements for the accuracy, precision, and detection limits for measurements of selected radionuclides in the bodies of, or in biological samples from, persons occupationally exposed to the intake of radioactive materials. It also provides standard quality control procedures for the internal quality assurance programs of radiobioassay laboratories, and criteria to be used as a basis by testing laboratories to evaluate the conformance of radiobioassay service laboratories to both the quantitative performance criteria for accuracy, precision and detection limits, and standard quality control procedures.
Besides ongoing developments in the dosimetry of incorporated radionuclides, there are various efforts to improve the monitoring of workers for potential or real intakes of radionuclides. The disillusioning experience with numerous intercomparison projects identified substantial differences between national regulations, concepts, applied programmes and methods, and dose assessment procedures. Measured activities were not directly comparable because of significant differences between measuring frequencies and methods, but also results of case studies for dose assessments revealed differences of orders of magnitude. Besides the general common interest in reliable monitoring results, at least the cross-border activities of workers (e.g. nuclear power plant services) require consistent approaches and comparable results. The International Standardization Organization therefore initiated projects to standardise programmes for the monitoring of workers, the requirements for measuring laboratories and the processes for the quantitative evaluation of monitoring results in terms of internal assessed doses. The strength of the concepts applied by the international working group consists in a unified approach defining the requirements, databases and processes. This paper is intended to give a short introduction into the standardization project followed by a more detailed description of the dose assessment standard, which will be published in the very near future.
Basic differences in the methods used in ICRP Publications 2 and 30 to derive limits for the intakes of radionuclides by workers are discussed. Although the philosophy on dose limitation now adopted by the Commission has changed, this has resulted in only small changes in the limits on intake of radioactive materials. Much greater changes result from improved knowledge about radioactive decay schemes and the uptake and retention of radionuclides in body tissues. The limits on intake recommended in ICRP Publication 30, Part l, are about equally divided above or below the corresponding values given in ICRP Publication 2 and the geometric mean of all the ratios new/old is very close to unity, although in a few instances the values differ by an order of magnitude or more. (C)1981Health Physics Society