Invited Speakers

Abstract of Talk :

Since its foundation in 1928 the ICRP has continuously developed recommendations to contribute to an appropriate level of protection for people and the environment against the detrimental effects of radiation exposure without unduly limiting the desirable human actions that may be associated with such exposure.

The ICRP System of Radiological Protection is based on science, ethics, and experience.

Key principles are Justification (any decision that alters the radiation exposure situation should do more good than harm), Optimisation (doses should all be kept as low as reasonably achievable, taking into account economic and societal factor), and Dose Limitation (the total dose to any individual should not exceed the appropriate limits).

While the system is robust and applicable and used all over the world, it requires regular review to make sure that it is still up to date and fit for purpose.

Recently, ICRP has recently embarked on a review and revision of the System that will update the 2007 General Recommendations in ICRP Publication 103 from 2007.

This paper informs about the System and describes first steps taken by the ICRP towards the review and refinement of the System.

Abstract of Talk :

For the implementation of the radiation protection system, the concepts of risk tolerability and search for reasonableness were developed by the International Commission on Radiological Protection from the first publication in 1959. Since then, these concepts have been the subject of discussions and are at the core of the processes for evaluating protective actions. Thus, for a given exposure situation, it is necessary to ask whether the doses received or potentially received by the exposed people can be considered tolerable, with regard to the radiation-induced risk. In addition, with reference to the implementation of the principle of optimisation, it should be assessed whether all exposures are maintained or reduced as low as reasonably achievable taking into account economic and societal factors.

This approach has been largely developed in ICRP Publication 60. More recently, ICRP Publication 101 described the approach to address the implementation of the optimisation process including the involvement of stakeholders and further considerations to be taken into account to elucidate what is reasonably achievable.

With the Publication 138 on the ethical foundations of the system of radiological protection, specific attention has been devoted to the concepts of reasonableness and tolerability, calling for revisiting their meaning, role and practical implementation in the light of the core ethical values (beneficence and non-maleficence, prudence, justice and dignity) and procedural values (accountability, transparency and inclusiveness). In 2019, ICRP has set up a dedicated task group (TG114) to review the historical and current perspectives on reasonableness and tolerability in order to consolidate and clarify Publication 103, and to prepare the considerations and basis needed for development of future recommendations.

This presentation will address the following issues currently under investigation by the ICRP TG114:

• Rationale for the application of the tolerability of risk in different exposure situations as well as the borders with unacceptable level of risks;
• Comparison with the approaches adopted for managing other risks;
• Dynamics from tolerability to reasonableness in the process of implementation the radiological protection system;
• Importance of the deliberative process to be put in place with the stakeholders for evaluating “what is reasonable”, referring to good judgment, fairness, practicability and moderateness;
• Selection process for radiological criteria for different exposure situations and their link with dose limits and reference levels, relying on the radiological detriment as benchmark without limiting to numerical criteria.

Abstract of Talk :

Enhancing public understanding of radiation and risk is highlighted by experiences from past emergencies, including the accident at TEPCO's Fukushima Daiichi Nuclear Power Plant in 2011 and the following post disaster recovery, as one of the most important challenges, and this challenge is common across almost all public interfaces regarding radiation and risk.

There is a growing need and interest for the IRPA Associate Societies (AS) to enhance their programmes in this important area. This is a key but challenging activity which needs further support. To this end IRPA has been continuing a Task Group (TG) activity for Public Understanding on radiation and risk since 2013. In this presentation, recent TG activities for the key tasks will be reported.

Abstract of Talk :

Nuclear Power Corporation of India Limited (NPCIL) is currently operating 22 reactors in the country, including 2 Boiling Water Reactors (BWRs), 18 Pressurized Heavy Water Reactors (PHWRs) and 2 Pressurized Water Reactors (PWRs), with a total installed capacity of 6780 MWe. NPCIL has registered the highest-ever power generation of 47112 MU in the financial year 2021-22. NPCIL has over 582 reactor years of operating experience with no event of concern in public domain. Radiation Protection aspects in Nuclear Power Plants are being improved constantly through Design, Construction, Commissioning and Operation, which has resulted in improvement in collective radiation exposures (CRE). The collective dose profile over the four decades starting from 1969 to 2022 is shown in Figure-1.

The collective dose per Reactor per year has reduced to 0.64 person-Sv in the year 2022 in NPCIL. This could be achieved by improvements in the design and adopting good industrial operating practices. The major changes adopted in design include selection of materials with low Cobalt impurity, relocation of equipment, provision of adequate shielding in the design stage based on operating experience, Characterization of sources to identify the origin / reasons, Dissemination of lessons learned from National and International incidents occurred in NPPs through Experience Sharing Reports from Head Quarters, Development of Radiological Safety Performance Indicators to assess the Performance of NPPs exclusively in the area of Radiation protection and development of benchmarking based on RSPI evaluation for continual improvement.

Periodical evaluation of performance of NPPs in radiation safety aspects and addressing of potential areas identified for improvement by various audits/ peer reviews, by the NPCIL management in time resulted in achieving reduction in collective dose for plant personnel and members of the Public. Dose to radiation workers and members of the public are well within the AERB authorized limits.

The constant monitoring, analysis and implementation of appropriate design modifications has also resulted in reduction of radioactive effluents discharges from the NPPs. The estimated dose to a hypothetical person staying at the exclusion boundary of NPPs over the last five years was in the range of 0.2 to 40 µSv/year.

In addition, the insights provided during review process by regulators also motivate the utilities to achieve excellence in radiation safety performance.

NPCIL is committed to operate NPPs in a safe, environmentally friendly and an economically viable source of electrical energy by enhancing design modifications from time to time based on the operating experience of NPPs in India as well as the world over.

Abstract of Talk :

The presentation will give an overview of the recent Scientific Committee's published, ongoing and planned evaluations. It will focus on the new "Strategy to Improve Collection Analysis and Dissemination of Data on Radiation Exposure" adopted by the sixty-nineth session, and summarized the challenges and planned actions on the collection and analysis of data on public, medical and occupational exposures and enhanced collaboration with international organizations.

Abstract of Talk :

In order to promote safety improvements in radiotherapy, the IAEA developed an Internet-based voluntary anonymized non-punitive educational safety reporting and learning system for radiotherapy called SAFRON (SAFety in Radiation ONcology). The system was launched in 2012, aiming to enable global sharing of information on safety related events and safety analysis in order to improve the safe planning and delivery of radiotherapy. Over the years, it has been expanded to include not only external beam radiotherapy, but also brachytherapy and radionuclide therapy incidents and near-misses. In 2022, SAFRON was updated and migrated to a new platform, to enable multilingual support, mobile phone access, analysis capabilities and other features.

SAFRON currently contains more than 1700 reports of incidents from around the world. It allows the health professional to search for information that may be of value for enabling the safe use of radiation in radiotherapy centres. The type of information often includes not only the summarized case-history of the incident, but also who discovered the incident and how it was discovered, as well as causes and contributing factors to incidents, looking for important patterns in the processes. Furthermore, information on the safety barriers (processes or equipment) that worked or didn’t work in stopping or mitigating the event is noted for many events, along with preventive actions suggested by the event-reporter.

It is believed that the updated SAFRON system has the potential to enable medical facilities to use radiation theraputically in a safer way, when properly utilized.

Abstract of Talk :

Proton beam Therapy (PBT) is increasingly adopted as a superior radiotherapy technique owing to its Bragg Peak characteristics, which allows sparing of surrounding normal tissue more effectively than modern photon radiotherapy techniques. PBT can potentially enhance clinical outcomes, quality of life, and prevent/reduce the radiation induced secondary cancers. However, only 2-3% of radiotherapy patients, currently receive PBT due to limited proton therapy (PT) facilities and cost-benefit justification. Nevertheless, over the last decade, the number of PT facilities have increased significantly, primarily due the technological development which have focussed on making cost effective, compact (single-room) PT facility rather than usual capital-intensive multi-room facility, to serve a broader group of oncology patients. Currently, more than 110 PT facilities are operational across the world and many more are under construction and planning stage. Apollo Proton cancer centre is the first multi-room PT facility in South-East Asia and Middle East which was clinically commissioned in Jan 2019.

In PBT, a narrow proton beam (70-250 MeV) accelerated either by cyclotron or synchrotron, is transported through a beam-line to the treatment room. Depending on the design of the “nozzle”, last part of the beam-line just before the beam enters the patient, a therapeutic dose can be delivered precisely to the tumour by spreading out the Bragg peaks of multiple proton energies either by using double scattering (DS) or pencil-beam scanning (PBS) techniques. During the process, proton interact with various materials while accelerating inside the cyclotron, transferring through the beam-line, treatment nozzle, patient specific beam shaping and range modulating devices and finally to patient. As a consequence, secondary stray radiations predominantly prompt neutron having wide energy spectrum ranging from thermal up to 230-250 MeV and photon up to ≈ 10 MeV are produced through intra nuclear cascade and evaporation. This causes a concern for radiation protection of staff and public.

A massive shielding barrier is therefore required to stop secondary radiations, especially high energy neutron for protection of occupational workers and general public from unwanted radiation. The shielding thickness can be calculated using analytical methods or more accurately using Monte Carlo (MC) codes [FLUCKA, MCNPX, GEANT 4] and computational models. However, designing the neutron radiation shielding for a PT facility presents several major challenges due to the uncertainties in the estimation of proton beam lost, neutron production and attenuation calculations. Even with the best available predictive tools, it remains large and difficult to estimate. Therefore, a detail measurement of total ambient dose equivalent H*(10) need to be carried out around the PT facility so as to ensure that the dose limit of radiation professional and public are within the stipulated limit.

Unwanted irradiation of patient from stray radiation is another concern of radiation protection while treating a clinical indication and it depend on proton delivery technique. In DS technique, beside two metal scatterers and range modulator present in nozzle, patient specific beam shaping devices made up of brass and 3D Tissue compensators are used resulting in production of secondary radiation sources in close proximity to the patient. In PBS technique, all these devices are replaced by a pair of scanning magnets resulting in a considerably lower stray radiation (neutron) compared with DS.

PBT of a patient entails a carefully orchestrated series of steps that include the following: a) preparation of customized patient immobilization device, b) 3D imaging (CT, MRI, PETCT), c) multi-modality image registration in the treatment planning system (TPS), d) delineation of tumour and critical organs in TPS, e) 3D modelling, simulation and dose computation of intended treatment plan using sophisticated dose calculation algorithms, f) seamless transfer of all treatment planning parameters from TPS to treatment delivery machine through oncology information system (OIS), g) assurance that planed dose in TPS is delivered in treatment machine, h) target localization on treatment delivery machine using on-bord imaging and i) treatment delivery. Any weak link in any of the stage may lead to compromise in treatment outcome and hence stringent quality assurance (QA) program must be design and implemented to guarantee that the intended dose is exclusively given to the tumour while maintaining the dose to the surrounding critical organs. The QA program must ensure that all the a) electro-mechanical, b) safety interlock, c) proton beam characteristics, d) integration of multi-modality equipment are functional and within their tolerance limit. In this talk, every facet of radiation safety concerning to general public, professionals, and patients will be discussed in a proton therapy facility.

Abstract of Talk :

The Nuclear Instrumentation in DAE is rooted way back in 1956 with the commissioning of APSARA reactor with home-grown knowledge. The visionary initiatives of Dr Homi J. Bhabha nucleated host of developments in electronics with fledging nuclear program in early decades that spawned a vibrant nuclear instrumentation and electronics community. Variety and complex nuclear electronics and instrumentation systems were developed for the expanding portfolios of nuclear plants – Apsara, Uranium Metal Plant, Nuclear Fuel Facility, Zerlina, Plutonium Plant, CIRUS, Tarapur Atomic Power Station (TAPS).

It was an exhilarating decade of innovation with a world class product varieties and R&D. The product portfolio was bursting at the seams, spilling into medical, industrial and other applications. This period also gave birth to Electronics Corporation of India Limited (ECIL) in 1967 for technology transfer and product assimilation. In subsequent decade the microelectronics grew in breakneck speed obsoleting itself every few years and eventually opening the sector for global competition in 1990’s led to the gradual decline and near demise of the domestic electronics industry and expanding Indian market survived mostly on imports. However, matters stood differently in DAE. After the events at Pokhran in 1974, DAE came increasingly under sanctions. Over the years, technology denial regimes became pernicious and all encompassing. DAE needed to continue to be self-reliant.

Embargos stimulated developments in nuclear electronics, C&I and computers ensuring that DAE programmes became largely invulnerable to sanctions. Thereby, rooted in-house knowledge base in nuclear electronics and systems faced with the circumstantial demands to grow the nuclear program in the country kept fuelling the activities in the department. This led to inhouse development of most of the core nuclear detector and instrumentation needs of research reactors, PHWRs (220Mwe, 540Mwe and now 700Mwe), upcoming PFBR, cPWR (Project B) and host of subsequent development activities of upgradation and indigenous substitute. It may be broadly stated that the country is largely self-sufficient in reactor nuclear instrumentation systems.

There has been progressive indigenous technological growth to meet the needs of plantwide radiation instruments and systems. The sustenance of most of these technologies through industries remains to be demanding. However, the R&D centres continue to work on the indigenous technological needs of such instrumentation for reprocessing plants of PHWR as well as FBR, safe-guarded plants besides addressing the security sector, industrial and medical including for mega science programs. The talk is an attempt to touch upon a holistic view of ecosystem that evolved into present status of self-sufficiency due to the visionary initiatives of Dr.H.J.Bhabha and maturity levels that possibly exists in this vital field of nuclear instrumentation.

Abstract of Talk :

The Japan Atomic Energy Agency has been developing a code for internal dose calculation in line with the 2007 Recommendations of the International Commission on Radiological Protection (ICRP). The code named as Internal Dose Calculation Code (IDCC) has two major functions: calculation of dose coefficients and estimation of intakes of radionuclides from monitoring data.

The function calculating dose coefficients is developed mainly to revise the Japanese regulatory standards for internal exposure. The existing standards are established as derived air concentrations for workers and concentrations of exhausted air and drain water from facilities for the public, which are determined from the effective dose coefficients based on the 1990 Recommendations of the ICRP for each radionuclide and its chemical form. Although the ICRP does not provide dose coefficients for radionuclides with half-lives of less than 10 minutes, the regulatory standards include values for those radionuclides. Therefore, we need a code to calculate dose coefficients in line with the ICRP 2007 Recommendations. The developed function was verified by confirming that the generated dose coefficients coincide with the values recorded in OIR (Occupational Intakes of Radionuclides) Data Viewer of the ICRP.

A dosimetry tool in conformity with the ICRP 2007 Recommendations is necessary for radiation safety management and for retrospective assessment in an intake incident. Then, we have also developed a function estimating intakes of radionuclides by using individual monitoring data obtained from in-vivo measurement and bioassay. This function is applicable to the conditions of a single acute intake, multiple acute intakes, and chronic intakes. The maximum likelihood method is applied to estimate intakes in fitting the time course of the activities in the body and organs and/or excretion rate generated by the code to monitoring data. The function is verified by confirming that the estimated intakes coincide with the literature values for case studies of IMBA code and the IDEAS guidelines.

The function estimating intakes will be used not only by researchers and technical staffs of radiation protection but also local government staffs concerning nuclear emergency preparedness. Therefore, usability is important. Then, we developed a graphical user interface (GUI) to integrate the two functions and to operate them. Figure 1 shows samples of the windows of the GUI of the IDCC. Users can select calculation modes from calculating dose coefficients or estimating intakes, and can set calculation conditions instinctively by using pull-down menus. The GUI also enables to edit data of biokinetic models and specific absorbed fractions (SAF). The data editing capability is useful for individual retrospective dose assessment. In addition, installers for Windows, Mac and Linux PCs were developed for ease of use. Development of the arithmetical operation part of IDCC has been completed. Now, we are working to incorporate the biokinetic models of ICRP Publ. 151. In the future, we plan to implement data for members of the public and SAF data for the average adult Japanese to IDCC. In the presentation, we will introduce the overview of the IDCC and the distribution of the IDCC.

Abstract of Talk :

"Tailings" are the chemically modified by-products generated in bulk quantities from the ore processing facilities in uranium mining and milling industry. It is produced during the ectraction process of uranium which is the nuclear fuel of the current nuclear power programme of the country. As per current practices, large quantities of uranium mill tailings are stored in engineering impoundments, called Tailings Pond, for long-term in order to exercise proper control on the source-term. The concern arises because of the very large amounts of tailings, associated with naturally occurring radionuclides and residual uranium, requiring recycling or safe disposal. Bulk quantities of tailings covering large land areas are of a great concern in the long term, specifically after facility operation for its structural integrity, monitoring requirements and regulatory compliance. Nationally under the focal theme of circular economy, scientific efforts are made to utilise the waste and recover the leftover economically valuable elements and the philosophy has been adopted in many NORM industries like utilization of fly ash and phosphogypsum.

There is a general recognition that at some level, the society is now expecting practices that put less pressure on natural resources and entail the recycling and reuse of materials. In view of the presented data, the Indian U mill tailings can be used as building materials in different construction activities. The classification system based on Radium equivalent is also followed in Austria, Finland, Slovakia, Latvia, Norway, Israel and Russia. Preliminary studies, both experimental and theoretical, were carried out to explore the feasibility of plausible utilization of uranium mill tailings in construction industry within the regulatory framework.

Utilisation of uranium mill tailings generated from the Turamdih process plant was studied by making bricks. Tailings samples were collected from the Turamdih tailings pond and were used to make bricks by moulding and pressing techniques. In the range of 15–62% mill tailings along with other materials such as sand, cement, gypsum, lime, bed materials, LD (steel) slag, fly ash, and aggregates were used to make bricks to understand its utilization as building material. Molded bricks and pressed bricks with different composition have been prepared for the experiments. Compressive strength, water absorption, leaching test and 222Rn mass exhalation rate were studied using standard methods.

The major findings were discussed in the present paper. The preliminary results are encouraging and support scientific intent for utilization of uranium mill tailings for various societal applications.

Abstract of Talk :

In this study, a Monte Carlo (MC) calculation code was employed to investigate homogeneous and/or inhomogeneous exposure scenarios with reference to actual radiation accidents (IAEA 2002, and IAEA 2019) and to evaluate the relationship between readings from personal dosemeters, organ doses, and computed effective doses. A Monte Carlo calculation code PHITS 3.24 was employed for a series of calculations (Sato et al., 2018). The International Commission on Radiological Protection (ICRP) computational adult male phantom was introduced as a radiation worker (ICRP, 2007, ICRP 2010). As exposure scenarios, in addition to the ICRP-specified irradiation condition of anterior-posterior (AP) irradiation geometry, point radiation source was set on the right chest or on the left hip to mimic inhomogeneous exposure accidents may be encountered in the industrial radiography.

Abstract of Talk :

The radiation protection program for industrial and medical uses of radiation in India had its origin in the nuclear emulsion activities of TIFR for cosmic ray research. The radiation monitoring program was initiated in Atomic Energy Establishment (AEET) at TIFR in 1952 for a small group of radiographers working with X-rays in hospitals using Kodak photographic film as a film badge system. This film badge service established in AEET later developed into a countrywide personnel monitoring programme. The story of transition of personnel dosimeter based on Kodak NTA films to thermoluminescence dosimeter (TLD), a new form of dosimeter with dose linearity over six decades, is one of the most successful examples of indigenization in any area in India. Soon, the instrumentation to study TL phosphor was developed and a large team gravitated to R&D in TL materials. Innovative methods to make a coil heater to heat the TL material on a Kanthal strip with a thermal shield resulted in a TL reader which was constantly improved upon from time to time. Initially, the TL glow curve was recorded on a strip chart recorder. Later, standalone TL readers for research applications were developed. Large scale applications such as studying the biological effects of the natural high background radiation areas (HBRAs) of Kerala on the population, and national radiation map project by deploying TLDs throughout the country are the important applications.

During the 1970s, CaSO4:Dy Teflon-based TLD badge was first introduced for personnel monitoring service at the Tarapur Atomic Power Station using an indigenously developed manual TLD badge reader. Based on the superior performance of the TLD badge system, the film badge service was gradually replaced with the TLD badge system in the Department of Atomic Energy (DAE) by 1986 and in non-DAE institutions by 1998. From 2000, the manual TLD badge readers were completely replaced by PC-based, semiautomatic TLD badge readers with several performance improvements. Maintenance of occupational dose records of all the radiation workers was also initiated with the inception of the personnel monitoring services around 1953. In the last decade, extensive efforts were made to develop indigenous optically stimulated luminescence (OSL) phosphors and dosimetry systems for personnel and environmental monitoring as an alternative to the TLD-based system. OSL based whole body and eye lens badges were developed based on α-Al2O3:C phosphor.

Internal dosimetry, a branch of radiation protection, deals with the measurement of internal contamination and assessment of internal dose. The Health Physics Division in the then AEET started internal dosimetry activities in 1957 by setting up a Bioassay laboratory in the TIFR campus. The development of a methodology for monitoring tritium in environmental and bioassay samples, was started and established using the gas-phase counting technique. The Bioassay Group developed methodologies for the estimation of actinides and fission products in urine samples. The procedures were further updated in the later years for ultra-trace level detection of radionuclides, especially actinides in biological samples. A separate Body Burden Measurement Section was formed in HPD while the CIRUS reactor was commissioned for whole-body monitoring of workers during 1960. NaI(Tl) detector based Whole Body Monitor (WBM) was developed inside a 4 ton steel shield room (Chhabra, 1964). A press wood phantom simulating human body torso containing 137Cs and 60Co standard sources was fabricated for calibration purpose. Later, a 200-ton special virgin steel room (free from radioactive contamination) was fabricated at BARC Hospital basement for measurement of low energy photon emitters (< 200 keV) like 239Pu (13-20 keV), 241Am (60 keV), 238U (63, 93 keV) etc. Special detectors like Phoswich (Sharma et al., 1975) and an array of HPGe detectors along with PSD are installed inside these steel rooms. After the nuclear accident at Chernobyl, BOMAB phantoms (IAEA, 2007) of different age groups 1-, 5-, 10-, 15- and 20 years were fabricated for calibration of WBM to meet the requirement of population monitoring. Field deployable portable in-vivo monitoring systems were developed for emergency response. Also, a Quick Scan -WBM was developed for rapid monitoring with counting time of as low as 1 min. The talk covers the historical developments of dosimetry practices in India.

Abstract of Talk :

Aerial Gamma Spectrometry is an effective tool for rapid assessment of radiological impact over a large area following a nuclear or radiological emergency. Aerial Gamma Spectrometry System (AGSS) is deployed on a helicopter or a fixed wing aircraft to carry out radiation monitoring of the affected area. The system has capability to generate radiation profile of the surveyed area based on the level of activity deposited on the ground due to fallout. The development and application of Aerial Gamma Spectrometry Systems has been reported by different institutions and organizations. Radiation Safety Systems Division of Bhabha Atomic Research Centre has developed an Aerial Gamma Spectrometry System (AGSS) which has the capability of quantitative and qualitative (identification of radionuclide) assessment. The system consists of a 3" × 3" NaI(Tl) Scintillation Detector, a Pulse Height Analyzer built around an industrial workstation and a Global Positioning System receiver to obtain positional coordinates. Spectral data is acquired by the system at a preset interval, generally every two seconds, and is tagged with the corresponding positional information. Extent of radioactivity deposited on ground is computed on-line based on the altitude of flying (generally, 80 m to 150 m) by using Air to Ground Correlation Factors which have been established through calibration exercises carried out over simulated ground contamination. The on-line mapping enables easy demarcation of the contaminated area based on the level of radioactivity and any hot spot on ground also can be identified. The minimum detectable activity of AGSS is 25 kBq.m-2 (90 nGy.h-1 at 1 m above ground) for 137Cs from an altitude of 100 m. Several aerial radiation monitoring exercises have been carried out using AGSS to demonstrate the has the capability of the system for identification of radionuclides and detection of ground contamination. The system can also be effectively used to search and locate orphan radioactive sources.

Abstract of Talk :

Special Nuclear Material (SNM) detection and quantification assumes importance in alpha waste from nuclear fuel cycle facilities. The present talk dwell on various techniques currently in use at IGCAR for waste assay. Delayed neutron counting method is the conventional method employed for this purpose which is based on the measurement of spontaneous neutron emission of 240Pu. However, the presence of other actinides in the waste would complicate the counting which results into errors in waste assay. Use of neutron coincidence technique in which it will be solved by differentiating the neutron originating from 240Pu and other actinides is discussed.

This talk will cover the design and development aspects of system which is under expansion at SQRMG, IGCAR, Kalpakkam.

Abstract of Talk :

After the Fukushima Daiichi Nuclear Power Plant accident, external exposure from radionuclides deposited on the ground is by far dominant pathway contributing effective dose in the most affected regions. In 2013, the Nuclear Regulation Authority of Japan stated the importance of residents understanding their own exposure doses. Personal dosimeters are generally used to measure individual exposure doses; however, they cannot be used for predictive or retrospective evaluations and it can be burdensome to wear personal dosimeters at all times. Therefore, a model was developed to estimate exposure doses on the basis of air dose rates and personal behavior (Sato et al (2019)). The model can predictively and retrospectively estimate realistic individual effective doses without burden by changing the air dose rate data used for estimation. We have developed three types of external exposure dose simulators based on the model to meet the need for residents to more easily estimate their own doses. By utilizing these simulators according to the purpose, residents can easily grasp their own exposure doses. Currently, we are lending the devices on which developed simulators are installed to municipalities with specified reconstruction and revitalization base in Fukushima Prefecture, and the simulators are used as one of risk communication tools. The web-based simulator is being prepared to be available as open source software from Japan Atomic Energy Agency. In addition, we are considering linking our simulators with the database for radioactive substance monitoring data (https://emdb.jaea.go.jp/emdb/) for greater convenience in the future. The developed simulators can be used in case of future nuclear disaster by adjusting the calculation parameters.

Abstract of Talk :

Over many years, the IAEA programmes in environmental modelling and dose assessment have provided a platform for exchanging information in radiological environmental impact assessment (REIA), contributing to capacity building and knowledge management in the field. In the latest in this series of IAEA initiatives, the IAEA MEREIA Programme, this component is emphasized, with a dedicated agenda for developing young and less experienced professionals. Launched in 2021 in a virtual technical meeting, MEREIA offered five webinars on basic radiological impact assessment concepts over the past year, each attended by more than 100 young professionals. The 2nd MEREIA technical meeting, a hybrid event held from 28 November to 2 December, brought together 90 in-person and 150 online participants from different backgrounds. During the week, the young professionals had the opportunity to engage in working group activities and benefit from interacting with senior professionals while developing strategies to assess radiological impact in different scenarios based on real situations. The meeting also offered morning training sessions and featured mentoring activities besides the regular programme work. The MEREIA Mentoring Scheme was officially launched and will allow young professionals to access more readily the vast pool of experienced professionals within the MEREIA community.

The way forward for MEREIA activities is to connect and inspire professionals worldwide to work collaboratively to enhance capability and capacity in radiological environmental impact assessment (REIA) within the context of broader processes for environmental impact assessments (EIA). The programme is open to professionals from regulatory bodies, operating organizations, scientists, researchers, decision makers and others with responsibilities in radiological environmental impact assessments.

Abstract of Talk :

In 2019, IRPA formed a task group dedicated to radiation protection issues associated with naturally occurring radioactive materials (NORM) in industry. The task group is made up of recognized radiation protection practitioners from a number of countries and provides a unique practical perspective on management and regulation of NORM. The task group meets regularly and discusses existing and emerging NORM related issues. An objective of the task group as defined by the IRPA Executive in original terms of reference, was to develop best practice guidance. This has evolved and the primary focus of the task group is to develop a practitioner’s handbook on good practices in NORM with the intention is to release this at the IRPA Congress in Orlando in 2024. Other activities of the task group include the regular participation in international conferences and forums. This paper will provide an overview and update on the work and activities of the task group including a description of an upcoming publication related to good practice in industries with NORM.

Abstract of Talk :

The protection of general public and the environment at large around a nuclear facility is achieved by following compliance to rigorous standards stipulated by regulatory agencies. This is achieved both at the level of design and at the level of operation through comprehensive research, development and application-based studies on: Environmental surveillance and radiation protection; Environmental impact, dose and risk assessment for radiological and chemical contaminants; Development of methodologies for determination of radionuclides, present at very low levels in various environmental matrices etc. The talk presents the various environmental and radiological surveillance practised at the operating nuclear power plant sites in India and the Indian Environmental Radiation Monitoring Network (IERMON) which is responsible for the realtime environmental radiation monitoring across the country.

Abstract of Talk :

Radiation dose projections involving atmospheric dispersion and deposition calculations are highly needed for initiating protective actions in the event of accidental releases from nuclear power plants. Ingestion of contaminated food is one of the major pathways of radiation exposure during nuclear emergencies. Estimation of ingestion dose is complex due to uncertainties in the information on food supplies in the emergency domain. Information on the crop lands, pasture lands etc. is needed to obtain reliable estimates of ingestion dose. The talk illustrates a method based Remote sensing derived Land Use/ Land Cover data for spatial ingestion dose estimation. This scheme based on the IAEA method is incorporated in the ONERS decision support system.

Abstract of Talk :

Radioactive substances and Radiation generating equipment have many beneficial applications, ranging from power generation to uses in medicine, industry and agriculture. Radiation sources, if not properly controlled or regulated can pose radiation risks to workers, public and to the environment. This necessitates appropriate mechanism for safety management and a regulatory framework to exercise control over the safe usage of the radioactive materials and the radiation generating equipment.

The regulatory control provisions shall ensure that the facilities engaged in applications of radiation (e.g. radiotherapy, nuclear medicine, diagnostic radiology, industrial radiography, radiation processing plants, nucleonic control systems etc.) do not cause undue risk to the health of occupational worker, general public and the environment. Regulatory control measures need to establish procedures for safe handling, storage, transportation and disposal of radiation sources/ equipment. The regulatory framework for exercising control over radiation sources has been well established in our country. Atomic Energy Regulatory Board (AERB), the national regulatory body of our country is empowered to regulate the radiation facilities under the Atomic Energy Act-1962. AERB develops and lays down safety requirements and guidance for facilities in the form of Regulatory Safety Documents. These documents comprise Safety Codes, Standards, Guides and Manuals. In order to carry out regulatory process of radiation facilities effectively and efficiently, AERB has developed an online system e-LORA (electronic Licensing Of Radiation Applications) wherein, applications can be submitted for various regulatory consents. This is an integrated system with emphasis on tracking down radiation sources starting from the procurement to its safe disposal. The safety review and processing of applications through the e-LORA system, facilitates the radiation facilities to follow the status of their applications and also helps AERB to keep the inventory of the radiation sources being used in the country. Further, keeping in mind the paramount importance of radiation safety, the regulatory inspections to verify the regulatory compliance in radiation facilities are also carried out periodically by AERB. The inspections are based on the established procedure which have been developed considering the hazard potential associated with the radiation installations.

The high standard of safety in multifarious applications of radiation can be achieved through the inherent built-in safety features incorporated in the design as well as operational and administrative control measures. The prime responsibility for ensuring safety and security of radiation sources, though rests with the operating organization, the real challenge always lies in minimizing the risk and utilizing the enormous technological potential for achieving a better quality of life for humankind. In the recent past, Government of India invited Integrated Regulatory Review Service (IRRS) team of IAEA to peer review the regulatory framework that also included the ‘Radiation Source Facilities and Activities’. After the peer review, IRRS team concluded that AERB showed a strong commitment and professionalism to ensure nuclear and radiation safety in the country and generally implements the regulatory process and safety requirements in accordance with the IAEA safety standards