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).