Health effects of security scanners for passenger screening, based on X-ray technology

Due to increased concern over terrorist attacks on aircraft, new technologies have been developed to improve the efficiency of security screening of passengers. Some of these technologies use ionising radiation such as X-rays for screening the passengers.  European Union formed a scientific committee to assess the risks related to use of security scanners for passenger screening that use ionising radiation. The X-ray based security screening technology used for the purpose relies on two techniques: backscatter or transmission. In the backscatter technique, radiation is reflected from the subject and detected to form an image of the body showing any concealed objects worn on the body. The transmission technique detects X-rays emitted by the equipment that pass through the body of the subject. Any concealed object provides an image by attenuating the radiation. While the backscatter technique can only reveal objects at the surface of the body, the transmission technique also shows objects within the body if their contrast differs sufficiently from the surrounding body fluids or tissue.

The health risk from the exposure is quantified in terms of effective dose, which takes into consideration the type of radiation and the sensitivity of the body parts exposed. The effective doses per scanned passenger are in the µSv range for the transmission technique and less than 1 µSv for the backscatter technique. The organ doses have generally the same order of magnitude. For persons scanned three times every working day, security scanning would result in an incremental effective dose of approximately 300 µSv (0.3 mSv) per year for the backscatter technique and close to 3 mSv per year for the transmission technique (assuming doses of 0.4 and 4 µSv per scan, respectively). The latter would exceed the dose limit of 1 mSv per year for the general public and hence would not comply with the current radiation protection standards for the most exposed group of frequent fliers. The risk of exposure using the backscatter technique can be considered as negligible.

Short-term (deterministic) health effects due to tissue damage cannot result from the doses delivered by security scanners. The long-term effects of ionising radiation include an increased cancer risk, which is assumed to be directly proportional to the dose received, without a safe threshold. However, direct evidence of an increased cancer risk or other stochastic risks in humans is not available at this low level exposure situations.

The potential magnitude of cancer risk from doses received from security scanners cannot be estimated, but is likely to remain so low that it cannot be distinguished from the effects of other exposures including both ionizing radiation from other sources (including natural) and background risk due to other factors. While the expected health detriment will probably be very close to zero for any single scanned person, the assessment of acceptability of the introduction of the security scanners using X-rays for passenger screening should also take into account the possible effect at the population level. Due to the substantial uncertainty regarding the potential occurrence of any health effects, risks for special groups within the population could not be evaluated meaningfully, although a higher risk related to exposure in childhood was noted (Source: Scientific Committee on Emerging and Newly Identified Health Risks, European Union, April, 2012) 

New ICRP Publication: Assessment of Radiation exposure of Astronauts in space; ICRP Publication No. 123, 2013

G. Dietze, D.T. Bartlett, D.A. Cool, F.A. Cucinotta, X. Jia, I.R. McAulay, M. Pelliccioni, V. Petrov, G, G. Reitz, T. Sato 

Astronauts, during their occupational activities in space, are exposed to ionising radiation from natural radiation sources present in the environment. The exposure assessment and risk-related approach described in this report is clearly restricted to the special situation in space, and should not be applied to the system of radiological protection followed on Earth.

The report describes the terms and methods used to assess the radiation exposure of astronauts, and provides data for the assessment of organ doses. Chapter 1 describes the specific situation of astronauts in space, and the differences in the radiation fields compared with those on Earth. In Chapter 2, the radiation fields in space are described in detail, including galactic cosmic radiation, radiation from the Sun and its special solar particle events, and the radiation belts surrounding the Earth. Chapter 3 deals with the quantities used in radiological protection, describing the Publication 103 system of dose quantities, and subsequently presenting the special approach for applications in space; due to the strong contribution of heavy ions in the radiation field, radiation weighting is based on the radiation quality factor, Q, instead of the radiation weighting factor, wR. In Chapter 4, the methods of fluence and dose measurement in space are described, including instrumentation for fluence measurements, radiation spectrometry, and area and individual monitoring. The use of biomarkers for the assessment of mission doses is also described. The methods of determining quantities describing the radiation fields within a spacecraft are given in Chapter 5. Radiation transport calculations are the most important tool. Some physical data used in radiation transport codes are presented, and the various codes used for calculations in high-energy radiation fields in space are described. Results of calculations and measurements of radiation fields in spacecraft are given. Some data for shielding possibilities are also presented. Chapter 6 addresses methods of determining mean absorbed doses and dose equivalents in organs and tissues of the human body. Calculated conversion coefficients of fluence to mean absorbed dose in an organ or tissue are given for heavy ions up to Z = 28 for energies from 10 MeV/u to 100 GeV/u. Doses in the body obtained by measurements are compared with results from calculations, and biodosimetric measurements for the assessment of mission doses are also presented. In Chapter 7, operational measures are considered for assessment of the exposure of astronauts during space missions. This includes pre-flight mission design, area and individual monitoring during flights in space, and dose recording. The importance of the magnitude of uncertainties in dose assessment is considered.

Annex A shows conversion coefficients and mean quality factors for protons, charged pions, neutrons, alpha particles, and heavy ions (2 < Z ≤ 28), and particle energies up to 100 GeV/u (Source: www.icrp.org). 

Latest ICRP Publication: Radiological Protection in Geological Disposal of Long-lived Solid Radioactive Waste, ICRP Publication 122, Ann. ICRP 42(3), 2013; W. Weiss, C-M. Larsson, C. McKenney, J-P. Minon, S. Mobbs, T. Schneider, H. Umeki, W. Hilden, C. Pescatore, M. Vesterlind

This report updates previous recommendations of the ICRP related to geological disposal of long-lived solid radioactive waste. The report explains how the ICRP system of radiological protection described in Publication 103 (2007) can be applied in the context of the geological disposal of long-lived solid radioactive waste. 

This report describes the different stages in the life time of a geological disposal facility, and addresses the application of relevant radiological protection principles for each stage depending on the various exposure situations that can be encountered. In particular, the crucial factor that influences the application of the protection system over the different phases in the life time of a disposal facility is the level of oversight or ‘watchful care’ that is present. The level of oversight affects the capability to control the source, i.e. the waste and the repository, and to avoid or reduce potential exposures. Three main time frames are considered: time of direct oversight, when the disposal facility is being implemented and is under active supervision; time of indirect oversight, when the disposal facility is sealed and oversight is being exercised by regulators or special administrative bodies or society at large to provide additional assurance on behalf of society; and time of no oversight, when oversight is no longer exercised in case memory of the disposal facility is lost (Source: www.icrp.org).

Assessing Radiation Dose of the Representative Person

Assessing Dose of the Representative Person for the Purpose of the Radiation Protection of the Public, ICRP Publication 101a, Ann. ICRP 36 (3), 2006: Dose to the public cannot be measured directly, and in some cases, it cannot be measured at all. Therefore, for the purpose of protection of the public, it is necessary to characterise an individual, either hypothetical or specific, whose dose can be used for determining compliance with the stipulated dose constraint/limits on individuals from specified sources.  

These dose constraints apply to actual or representative people who receive occupational, medical, and public exposures. This individual is defined as the ‘representative person’ by the ICRP. It is assumed that the Commission’s goal of protection of the public is achieved if the relevant dose constraint for this individual for a single source is met, and radiological protection is optimised. 

This report updates the previous guidance and the methods available for estimating annual dose to the public. In selecting characteristics of the representative person, three important concepts used are: reasonableness, sustainability, and homogeneity. Each concept is explained and examples are provided to illustrate their roles. Doses to the public are prospective (may occur in the future) or retrospective (occurred in the past). Prospective doses are for hypothetical individuals who may or may not exist in the future, while retrospective doses are generally calculated for specific individuals. 

Unlike in the earlier recommendations, the ICRP now recommends the use of only three age categories for estimating annual dose to the representative person for prospective assessments. These categories are 0–5 years (infant), 6–15 years (child), and 16–70 years (adult). For practical implementation of this recommendation, dose coefficients and habit data for a 1-year-old infant, a 10-year-old child, and an adult should be used to represent the three age categories. 

In a probabilistic assessment of dose, whether from a planned facility or an existing situation, the Commission recommends that the representative person should be defined such that the probability is less than about 5% that a person drawn at random from the population will receive a greater dose. If such an assessment indicates that a few tens of people or more could receive doses above the relevant constraint, the characteristics of these people need to be explored and actions to modify the exposure should be considered. The Commission recognises the role of stakeholders in improving the quality, understanding, and acceptability of the characteristics of the representative person and the resulting estimated dose (Extracted from the www.icrp.org).

Compendium of Dose Coefficients based on ICRP Publication 60, ICRP PUBLICATION - 119, Approved by the Commission in October 2011, Published by Elsevier Ltd., 2012, K. Eckerman, J. Harrison, H-G. Menzel and C.H. Clement

This report is a compilation of dose coefficients for intakes of radionuclides by workers and members of the public, and conversion coefficients for use in occupational radiological protection against external radiation from Publications 68 (1994), 72 (1996), and 74 (1996). It serves as a comprehensive reference for dose coefficients based on the primary radiation protection guidance given in the ICRP Publication - 60 recommendations (ICRP, 1991). 

The coefficients tabulated in this publication will be superseded in due course by values based on the ICRP Publication - 103 recommendations (ICRP, 2007). May be, it will be a long wait!

Radiation Safety of Gamma, Electron and X Ray Irradiation Facilities Specific Safety Guide, IAEA Safety Standards Series SSG-8 94 pp.; 8 figures; Language: English, Date Published: 2010

This Safety Guide provides recommendations on how to meet the requirements of the Basic Safety Standards with regard to irradiation facilities. It gives practical information on the safe design and operation of gamma, electron and X ray irradiators in accordance with these requirements, and discusses the beneficial applications of ionizing irradiation and how to avoid potential radiation hazards at industrial irradiators, including contamination arising from damaged radioactive sources. The Safety Guide is intended for use by the designers and operating organizations of these facilities and also by regulatory bodies. 

Contents: 1. Introduction; 2. Justification of practices; 3. Types of irradiator; 4. Principal elements of practices; 5. Individual monitoring of workers; 6. Workplace monitoring; 7. Control over radioactive sources; 8. Irradiator design; 9. Testing and maintenance of equipment; 10. Transport, loading and unloading of radioactive sources; 11. Emergency preparedness and response (Source: www.iaea.org).

Recommended dosage for iodine prophylaxis following nuclear accidents

As per the WHO guidelines (1999) for stable Iodine Prophylaxis, the recommended adult dose is 100 mg of iodine (130 mg of KI) for persons above 12 years. 

Children (3-12 y) - 50 mg of iodine (65 mg of KI)

Infants (1m to 3 months) - 25 mg (32 mg of KI)

Neonates (birth to 1 month) - 12. 5 mg (16 mg of KI)

To obtain full effectiveness of stable iodine for thyroidal blocking, it has to be administered shortly before exposure or as soon after as possible.This protective action is taken to prevent deterministic effects (Hypothyroidism) in the thyroid from the high levels (several Gy) of radiation dose to the thyroid from the uptake of radioiodines (mainly I-131, I-132 & I-133), released from the nuclear accidents, and to reduce the risk of stochastic effect - induction of thyroid cancer. 

Intakes can take place through ingestion/inhalation routes. The best estimate of excess absolute cancer risk is 4.4 × 10 to the power -4 per Gray per year for persons exposed before the age of 15, and virtually no risk is observed for exposure after the age of 40. 

The mass of thyroid varies with the age. Indian data (Source: Asian Reference Man Data, IAEA-TECDOC-1005) show variation from 1.5 gm (newborn), 8 gm (10 years) to 19 gm (male adults). Lower the mass, higher is the dose received for a given uptake and hence greater is the cancer risk.  

There is a greater need to protect the thyroid gland of the pregnant woman since the iodine uptake can be increased as compared to other adults. As much as 1/4 of the iodine taken by the mother may be secreted in the milk within 24 h. Newborn infants are quite likely the critical group of concern when deciding on the implementation of stable iodine prophylaxis. 

A generic intervention level of 100 mGy avertable dose is recommended for all age groups. However, the recommended intervention level for childhood exposure is 10 mGy avertable dose to the thyroid.

(Source: WHO Guidelines for Iodine Prophylaxis following Nuclear Accidents - Update 1999)

IAEA Publication: Storage of Spent Nuclear Fuel, IAEA Safety Standards Series SSG-15 Subject Classification: Radioactive waste management, STI/PUB/1503, 110 pp. Language: English, Date Published: 2012.

This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel.It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. Guidance is provided on all stages in the lifetime of a spent fuel storage facility, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. 

Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections in operating procedures for a spent fuel storage facility; Annex IV: Related publications in the IAEA Safety Standards Series; Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII:Postulated initiating events for consideration in a safety assessment (internal phenomena). Source: www.iaea.org.

IAEA publication: Evaluation of Seismic Safety for Existing Nuclear Installations Safety Guide, IAEA Safety Standards Series NS-G-2.13 84 pp, Language: English, Date Published: 2009

This Safety Guide provides recommendations regarding the criteria and methodologies to be used for seismic safety evaluation of existing nuclear installations, including installations whose purpose and associated radiological risks have changed, installations where longer term operation is under consideration and installations where comprehensive seismic safety reassessments have become necessary. 

Two methodologies are discussed in detail: deterministic seismic margin assessment (SMA) and seismic probabilistic safety assessment (SPSA). 

Contents: 1. Introduction; 2. Recommendations on formulation of the programme for seismic safety evaluation; 3. Data collection and investigations; 4. Assessment of seismic hazards; 5. Methodologies for the evaluation of seismic safety; 6. Nuclear installations other than power plants; 7. Considerations in upgrading; 8. Management system for seismic safety evaluation; Annex: Methodologies for seismic safety evaluation (Source: www.iaea.org).

Radiological Protection in Cardiology, ICRP Publication 120

Cardiac nuclear medicine, cardiac computed tomography (CT), interventional cardiology procedures, and electrophysiology procedures are increasingly used in medicine and form the major share of patient radiation exposure. Some of the percutaneous coronary interventions and cardiac electrophysiology procedures are associated with high radiation doses. These medical procedures can result in patient skin doses that are high enough to cause radiation injury and an increased risk of cancer. 

The Commission provided recommendations for radiological protection during fluoroscopically guided interventions in Publication 85, for radiological protection in CT in Publications 87 and 102, and for training in radiological protection in Publication 113. This report is focused specifically on cardiology, and brings together information relevant to cardiology from the Commission’s published documents. The material and recommendations in the current document have been updated to reflect the most recent recommendations of the Commission. 

This document provides guidance to assist the cardiologist with justification procedures and optimization of protection in cardiac CT studies, cardiac nuclear medicine studies, and fluoroscopically guided cardiac interventions. It includes discussions of the biological effects of radiation, principles of radiological protection, protection of staff during fluoroscopically guided interventions, radiological protection training, and establishment of a quality assurance programme for cardiac imaging and intervention (Source: www.icrp.org).