International Atomic Energy Agency Safety Research Articles

Evolution of zoning philosophy for emergency planning around nuclear power plants

Emergency planning zones (EPZs) serve as vital geographic framework within nuclear emergency preparedness and response (EPR) strategies, facilitating the swift implementation of protective measures across varying levels of urgency during nuclear emergencies. Over the years, nations have undergone thorough reassessments of these zones, in the light of evolving evaluation methods, reactor technology, and lessons learned from accidents, leading to the refinement of methodologies for delineating EPZ distances. Adherence to the International Atomic Energy Agency safety guidelines regarding EPZ size has demonstrated effectiveness in protecting both the public and environment surrounding existing nuclear power plants. As the landscape of nuclear energy evolves with the advent of small modular reactors, characterized by reduced power capacity and innovative safety features, there is enough potential for smaller EPZs. However, the determination of EPZ also depends on national policies, public acceptance, and nuanced risk criteria. This paper seeks to synthesize current insights into the evolution of EPZs and explores avenues for resizing them to accommodate emerging reactor technologies. Its findings hold relevance for reactor designers, risk analysts, and experts in nuclear EPR, offering guidance for future planning and implementation efforts.

A Source-Related Approach for Discussion on Using Radionuclide-Contaminated Materials in Post-accident Rehabilitation

Background: In the process of discussion on the possibility of using radionuclide-contaminated soil and debris generated by radiation disasters, a strategy for the proper management of radiation exposure protection while considering the source of the contaminated materials is necessary.Materials and Methods: The radiological protection criteria that are likely to be applied to the source-related approach based on the International Commission on Radiological Protection recommendations and the International Atomic Energy Agency safety standards are summarized. We proposed five interpretations of radiation protection to contribute to the promotion of discussion on the possibility of using a part of low-level-radionuclide-contaminated soil and debris in the post-accident rehabilitation. Interpretations I to III are based on the idea of “using a reference level to protect the public in post-accident rehabilitation,” whereas IV and V are based on the idea of “using the dose constraint to protect the public in the post-accident rehabilitation when the sources are handled in a planned activity.” The former idea is subdivided into three based on the definition of the source, which is managed by the reference level, and the latter idea is divided into two depending on whether or not additional dose from using contaminated materials is deemed acceptable.Results and Discussion: To confirm the applicability of the five interpretations presented, we suggested the concrete values of protection criteria via two feasible cases. In this case study, we proposed radiation protection by the dose constraint based on the Interpretation IV and chose 1 mSv/yr for the public and 20 mSv/yr for workers dealing with radionuclide-contaminated materials.Conclusion: We concretely and systematically demonstrated how the concept of radiation protection can be applied to the process of discussion on the possibility of using radionuclide-contaminated materials within the framework of an international system of protection. This study’s findings can provide necessary information to discuss the possibility of using radionuclide-contaminated materials as an alternative option for recovery and reconstruction after a radiation disaster from the viewpoint of radiation protection.

Assessment of doses in contaminated urban areas: modelling exercise based on Fukushima data

State-of-the-art dose assessment models were applied to estimate doses to the population in urban areas contaminated by the Fukushima Daiichi Nuclear Power Plant accident. Assessment results were compared among five models, and comparisons of model predictions with actual measurements were also made. Assessments were performed using both probabilistic and deterministic approaches. Predicted dose distributions for indoor and outdoor workers from a probabilistic approach were in good agreement with the actual measurements. In addition, when the models were applied to assess the doses to the representative person, based on a concept recommended by the International Commission on Radiological Protection and in the International Atomic Energy Agency Safety Standards, it was evident that doses to the representative person obtained with a deterministic approach were always higher than those obtained with a probabilistic approach using the same model.

Bounding uncertainties around the conceptual representation of species in radiological assessment in the context of routine atmospheric release

Wildlife protection has become of regulatory interest since the International Commission on Radiological Protection (ICRP) developed an approach to assess the level of radiological protection specifically for animals and plants. For the purpose of demonstrating compliance with regulation to protect the environment against routine authorised discharges from nuclear facilities, the wide variety of biota inhabiting an ecosystem needs to be condensed to a limited set of representative organisms, as proposed by the ICRP with a set of ‘reference animals and plants’ which can be considered representative of many other species. It is now recommended in the International Atomic Energy Agency Safety Standards, and internationally accepted, that the use of such a limited number of organisms to represent a pool of species is adequate for radiation protection purposes, particularly in planned exposure situations. Adding site-specific species to that set of surrogate species can respond to various interests, such as ensuring a site-specific context to the assessment that addresses stakeholder interests and can aid in stakeholder consultation and risk communication. Moreover, there is a need to question whether the use of the set of surrogate organisms is conservative enough to cover a wider range of biodiversity. Previous studies partially answered this question and this paper adds a range of test cases. A selection of hypothetical representations of possible site-specific species are assessed on the basis of possible variations in size (mass) and occupancy habits. Dose rates are evaluated to determine the greatest difference between hypothetical organisms and those for reference organisms (ROs), considering radionuclides (RNs) potentially discharged in atmospheric routine release from different nuclear facilities. Differences observed in the results between hypothetical organisms and ROs were less than one order of magnitude in all cases, the difference being dependent on the RNs considered. These findings do not preclude the inclusion of site-specific species in environmental radiological assessments if it is considered necessary, but they provide reassurance that using ROs for radiological impact assessments in the case of routine atmospheric discharges is sufficient.

Probabilistic Fault Displacement Hazard Assessment (PFDHA) for Nuclear Installations According to IAEA Safety Standards

ABSTRACT In the last 10 yr, the International Atomic Energy Agency (IAEA) revised its safety standards for site evaluations of nuclear installations in response to emerging fault displacement hazard evaluation practices developed in Member States. New amendments in the revised safety guidance (DS507) explicitly recommend fault displacement hazard assessment, including separate approaches for candidate new sites versus existing sites. If there is insufficient basis to conclusively determine that a fault is not capable of surface displacement at an existing site, then a probabilistic fault displacement hazard analysis (PFDHA) is recommended to better characterize the hazard. This new recommendation has generated the need for the IAEA to provide its Member States with guidance on performing PFDHA, including its formulation and implementation. This article provides an overview of current PFDHA state-of-practice for nuclear installations that is consistent with the new IAEA safety standards. We also summarize progress in an ongoing international PFDHA benchmark project that will ultimately provide technical guidance to Member States for conducting site-specific fault displacement hazard assessments.

Operational intervention levels for enabling the transition from an emergency exposure situation to an existing exposure situation following a radiological emergency involving release of radioactive material in the environment

Experience has demonstrated lack of preparedness at national levels to manage the consequences of a nuclear or radiological emergency in its later phase which, on occasions, resulted in unjustified actions. To assist Member States preparing for this phase, the International Atomic Energy Agency (IAEA) published guidance (IAEA Safety Standards Series No. GSG-11) recommending, inter alia, for operational intervention levels (OILT) to be developed for enabling the transition from an emergency exposure situation to an existing exposure situation. OILT are intended to also support decisions be made on lifting or adapting public protective actions imposed early in the emergency response and on actions to be taken to further reduce exposures. Using the methodology provided therein, OILT have been calculated for radiological emergencies involving release of the most commonly used alpha, beta and gamma emitting radionuclides in the environment. In addition, an approach for deriving a default OILT value has been presented and a default OILT value for the ambient dose equivalent rate at 1 m above ground level has been derived. The derived OILT values support the application of relevant IAEA safety standards (i.e. No. GSR Part 7 and No. GSG-11) as well as relevant recommendations of the Council Directive 2013/59 in relation to the transition from the emergency exposure situation to an existing exposure situation at national levels. However, they need to be integrated within the national protection strategies to guide the implementation of activities and actions that support the resumption of normal social and economic activity after the emergency which include those concerning return of members of the public, who were resettled during the implementation of evacuation or relocation, to their homes.

Best Practices in Physics-based Fault Rupture Models for Seismic Hazard Assessment of Nuclear Installations: Issues and Challenges Towards Full Seismic Risk Analysis

In recent years the International Atomic Energy Agency (IAEA) has been closely following and supporting the use of physics-based rupture models for ground motion prediction (e.g. IAEA Safety Standards Series No. SSG–9 and Safety Report Series No. 85) as well as for fault displacement prediction (IAEA-TECDOC, in preparation), respectively for applications in Seismic Hazard Analysis (SHA) and Fault Displacement Hazard Analysis (FDHA) for nuclear installations. Further strengthening of this effort and dissemination in practices for SHA, FDHA and engineering issues have been done through different international working group activities, being the most outstanding two international workshops on Best Practices in Physics-based Fault Rupture Models for Seismic Hazard Assessment of Nuclear Installations (BestPSHANI) in 2015 and 2018. A PAGEOPH topical volume for the BestPSHANI 2015 was published in Dalguer et al. (Pure Appl Geophys 174:3325–3329, 2017). Now, in this PAGEOPH topical volume we collect several articles from the BestPSHANI 2018 workshop as well as several new contributions. The issue also covers further topics on the assessments of engineering issues that rely on ground motion estimates for the evaluation of structures oriented to full seismic risk analysis. A total of twenty-nine papers have been selected covering topics ranging from the seismological aspects of earthquake source studies, ground motion and fault displacement modeling to the engineering application of simulated ground motion for the analysis of soil structure interaction, structural response and fragility curve analysis for the quantification of seismic vulnerability of structures and their seismic performance. Collectively, the seismological papers discuss several current issues of source characterization and ground motion prediction for SHA, highlighting the usefulness of physics-based models for future applications in practice. The engineering papers describe methodologies to develop integral models from source-to-structures that consider the developments of synthetic seismograms as input for structural response and fragility curves estimation for seismic vulnerability assessment. Therefore, this issue contents advanced seismological and engineering resources that might be useful to scientists, engineers, students and practitioners involved in all aspects of SHA, FDHA and vulnerability analysis of engineering structures for seismic risk.

Dominant factors influencing project quality in the radioactive minerals processing pilot plant construction

Monitoring and supervision of factors that predominantly affect the successful implementation of quality management systems on the project will improve the quality of construction that provides a safety factor in the construction of a pilot plant nuclear fuel processing, to determine those factors is the aim of this research. The separation of Uranium, Thorium, and Rare Earth Metals Pilot Plant was used as a case of the research. Adopted descriptive research design were the primary data collected using questionnaires. Data collected analysed thought multiple linear regressions that only done on factors that have correlation strong – very strong. International Standard Organisation series 9001:2015 and International Atomic Energy Agency safety standards series No. GS-R3 were used as independent variables (X), contamination level was used as dependent variable (Y) that reflected project quality. The result is there are three factors that statistically significant influencing project quality: leadership and commitment, project planning, and safety culture. Those factors influencing the achievement of construction quality of the pilot plant uranium, thorium and rare earth element i.e. 92.3%, while the remaining 7.7% (100-92.3%) is determined by other variables.

Investigation of Natural Radioactivity in a Monazite Processing Plant in Japan.

Monazite is a naturally occurring radioactive material that is processed for use in a variety of domestic applications. At present, there is little information available on potential radiation doses experienced by people working with monazite. The ambient dose rate and activity concentration of natural radionuclides in raw materials, products, and dust in work sites as well as the Rn and Rn concentrations in work sites were measured in a monazite processing plant in Japan. Dose estimations for plant workers were also conducted. The activity concentration of the U series in raw materials and products for the monazite processing plant was found to be higher than the relevant values described in the International Atomic Energy Agency Safety Standards. The ambient dose rates in the raw material yard were higher than those in other work sites. Moreover, the activity concentrations of dust in the milling site were higher than those in other work sites. The Rn concentrations in all work sites were almost the same as those in regular indoor environments in Japan. The Rn concentrations in all work sites were much higher than those in regular indoor environments in Japan. The maximum value of the effective dose for workers was 0.62 mSv y, which is lower than the reference level range (1-20 mSv y) for abnormally high levels of natural background radiation published in the International Commission of Radiological Protection Publication 103.

Safety assessment driving radioactive waste management solutions (sadrwms methodology) implemented in a software tool (safran)

In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 ( General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region.

Occupational Exposure to Natural Radiation in Zirconium Refractory Plants in Japan

The authors measured the ambient dose rate and activity concentration of natural radionuclides in raw materials, products, and aerosols on worksites, as well as the (222)Rn and (220)Rn concentrations in an unshaped refractory, a shaped refractory, and an electrocast refractory plant processing zirconium ore in Japan. Estimations were made of the effective doses to plant workers. The activity concentration of the (238)U series in raw materials and products in the refractory plants was higher than the critical values (10 Bq g(-1) for (40)K and 1 Bq g(-1) for all other radionuclides of natural origin) specified in the International Atomic Energy Agency Safety Guide. The ambient dose rate in the raw material warehouse of the electrocast refractory plant was 0.75 μSv h(-1), which was the highest among all the worksites at all the refractory plants studied. The activity concentrations of aerosols in the product-output site of the unshaped refractory plant was 0.0015 Bq m for U and 0.00078 Bq m(-3) for (232)Th, which were the highest of all the worksites for all refractory plants. The indoor (222)Rn and (220)Rn concentrations in all worksites of all the refractory plants were almost the same levels as those in everyday indoor places in Japan. The maximum value of the effective dose to workers was 430 μSv y(-1), which was lower than the intervention exemption level (1,000 μSv y(-1)) specified in ICRP Publication 82.

Safety and security during transport of industrial radioactive sources in Romania

The Romanian nuclear regulatory body National Commission for Nuclear Activities Control (CNCAN) takes seriously the challenge for users and regulators everywhere of ensuring a high level of safety and security in the use of nuclear materials and sealed radiation sources for peaceful purposes. Both Romanian and International Atomic Energy Agency safety standards are designed to facilitate the achievement of that goal. The Institute for Nuclear Research Pitesti manufactures 192Ir radioactive sealed sources for industrial gammagraphy that are transported by means of special designed packages under a special licensed car issued by CNCAN. Some aspects of safety and security during the manufacture and transport of 192Ir sealed radioactive sources are presented. A brief description of the packages used for the transport of iridium 192 sources is also given.

Methodological aspects for finite element modelling of lid systems for Type B(U) transport packages

The regulatory compliance of the containment system is of essential importance for the assessment process of Type B(U) transport packages. The requirements of the International Atomic Energy Agency safety standards for transport conditions imply high loading on the containment system. The integrity of the containment system has to be ensured in mechanical and thermal tests. The containment system of German spent nuclear fuel and high level waste transport packages usually includes bolted lids with metal gaskets. The finite element (FE) method is recommended for the analysis of lid systems according to the guideline BAM-GGR 012 for the assessment of bolted lid and trunnion systems. The FE analyses provide more accurate and detailed information about loading and deformation of such kind of structures. The results allow the strength assessment of the lid and bolts as well as the evaluation of relative displacements between the lid and the cask body in the area of the gasket groove. This paper discusses aspec...

Environmental assessment of advanced partitioning, transmutation, and disposal based on long-term risk-informed regulation: PyroGreen

The safety of spent nuclear fuel (SNF) storage has become one of the major issues of nuclear power plant operation. Direct disposal and recycling have both been criticized by the general public due to uncertainty of the long-term safety of SNF and high level waste (HLW) repositories. To meet the goal of sustainable nuclear energy, an innovative recycling approach using pyrochemical partitioning and transmutation termed PyroGreen, which is a conceptually advanced pyrochemical partitioning flowsheet, has been proposed to eliminate the need for HLW repositories. From the previous partitioning process, PyroGreen uses a combination of hull electrorefining, reductive extraction, and selective oxidation to further decontaminate SNF and HLW into low- and intermediate level waste (LILW). This paper examines the long-term environmental performance of a geological repository that houses all of the final PyroGreen-produced wastes while describing the feasibility of PyroGreen partitioning and transmutation. The final PyroGreen wastes are evaluated based on long-term risk-informed criteria: alpha-emitting isotope concentration, heat generation, and radiation dose in the surrounding biosphere. All final wastes arising from PyroGreen are to be disposed of in a geological repository at an intermediate depth, in compliance with new International Atomic Energy Agency Safety Guide for LILW. Migration assessment found that several long-lived fission products including C-14, Cl-36, Se-79, Sn-125, and I-129 dominate the long-term dose rate, whereas transuranic elements govern the risk of an inadvertent human intrusion scenario. In order to turn entire batches of SNF from 24 light water reactors with 1000 MWe with a design life of 40 years into LILW, we determined and compared the required decontamination factors of the key radionuclides for between intermediate-depth and near-surface disposal.

The new IAEA safety standards on probabilistic safety assessment

Abstract Increasingly wide application of Probabilistic Safety Assessment (PSA) to support a variety of issues dealing with operation and design of Nuclear Power Plants (NPP) imposes certain requirements on PSA quality in terms of its scope, level of detail, approaches, assumptions, and data used. If a unified set of balanced requirements is developed and consistently implemented, it would give confidence in using PSA results and insights in different applications and Risk-Informed Decision Making (RIDM); in addition, it would be possible to consistently compare PSA results for different plants. Such requirements are currently developed or being developed in several countries. The International Atomic Energy Agency (IAEA) is in the process of development of several publications in the IAEA Safety Standards Series that would provide guidance on the standard technical content of PSA. The paper discusses the latest developments at the IAEA in the area of PSA and provides a brief overview of the publications.

Experience with the operation, maintenance and utilisation of the 3 MW TRIGA Mark-II research reactor of Bangladesh

The 3 MW TRIGA (Training, Research, Isotope, General Atomics) Mark-II research reactor of the Bangladesh Atomic Energy Commission (BAEC) has been operating at Atomic Energy Research Establishment (AERE), Savar, Dhaka, since September 1986. Since its commissioning, the reactor has been used in various fields of research and utilisation, such as Neutron Activation Analysis (NAA), Neutron Radiography (NRG), Neutron Scattering (NS), manpower training and education, and production of radioisotopes for medical applications. The reactor facility encountered a couple of incidents, which were successfully handled by BAEC personnel. In some cases, the help of experts from various local organisations/institutions as well as from the International Atomic Energy Agency (IAEA) was obtained. The upgrading of the Safety Analysis Report (SAR) of the reactor facility was completed in 2005 as per the format of the IAEA Safety Guide, SG-35-G1. The cooling system of the reactor as well as some parts of the instrumentations used in the reactor systems were also upgraded/modified during this period. The paper highlights the experience with the operation, maintenance and utilisation of the research reactor for the last 21 years. It also presents some of the modification and upgrading works carried out to enhance the operational safety of the research reactor.

Issues on exemption levels for package surface contamination simply derived from IAEA TECDOC-1449 'Radiological aspects of non-fixed contamination of packages and conveyances'

AbstractThe present IAEA Transport Regulation on surface contamination (with limits specified in terms of Bq cm–2) is determined from a simple radiological model for the most hazardous radionuclides (Pu-239 for α emitters and Sr-90 for β emitters) and its extremely conservative model is applied for all other α and β emitters. In the International Atomic Energy Agency (IAEA) TECDOC-1449 report, the effect of radiation from non-fixed contamination on packages is evaluated and the dose conversion coefficients [(mSv/y)/(Bq/cm2)] are calculated for each radionuclide. In this study, exemption levels for surface contamination are calculated with the dose conversion coefficients specified in TECDOC-1449. The dose criterion for deriving the exemption level is chosen to be 0˙01 mSv/y, to maintain consistency with fundamental concepts adopted by the IAEA safety standards committees, namely, the Radiation Safety Standards Committee (RASSC), the Transport Safety Standards Committee (TRANSSC) and the Waste Safety Stand...