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[en] The CNSC (Canadian Nuclear Safety Commission) evaluates the safety performance of nuclear power plant (NPP) licensees and prepares an annual report on their safety performance referred to as the Regulatory Oversight Report, which is presented to the Commission and is subsequently published on the CNSC web page. Prior to 2017, the report was referred to as the Regulatory Oversight Report for Canadian NPPs. However, in 2017, the report was expanded to include the safety performance evaluation of waste management facilities located at NPP sites. The report has been renamed as the Regulatory Oversight Report for Canadian Nuclear Power Generating Sites. The CNSC evaluates how well licensees meet regulatory requirements and CNSC expectations for the performance of programmes in 14 safety and control areas (SCAs) that are grouped in accordance with their functional areas of management, facility and equipment, or core control processes. These SCAs are further divided into 71 specific areas that define the key components of the SCA. The functional areas, SCAs and the specific areas that are used in CNSC’s safety performance evaluation are presented. An example of safety performance ratings for Canadian NPPs is given. An example of a conclusion of a CNSC Regulatory Oversight Report for Canadian Nuclear Power Generating Sites is as follows: The evaluations of all findings for the SCAs show that, overall, NPP licensees made adequate provisions for the protection of health, safety and security of Canadians and the environment from the use of nuclear energy, and took the necessary measures to implement Canada’s international obligations.
[en] Seismic instrumentation systems are important elements for plant safety. They provide crucial information for the assessment of safety required to restart a plant after a shutdown caused by an earthquake. This publication presents information and experience related to the development of seismic instrumentation systems and the utilization of data recorded by them, based on recent seismic events. The publication is intended to be a reference for developing effective post-earthquake actions/procedures. It describes the typologies of seismic instrumentation systems and their application for obtaining ‘earthquake levels’ and ‘damage indicating parameters’ that are used to predict the extent of seismic damage caused by the recorded earthquake motions. It complements IAEA Safety Standards Series No. NS-G-1.6, Seismic Design and Qualification for Nuclear Power Plants, and Safety Reports Series No. 66, Earthquake Preparedness and Response for Nuclear Power Plants, as a technical publication relevant to seismic safety for new and existing nuclear installations.
[en] Why is leadership vital in nuclear safety? Leadership is needed to initiate appropriate safety actions, motivate staff to ensure safety procedures are followed 24/7, and provide guidance on implementing safety measures. Learning about the importance of leaders in safety is part of the IAEA International School of Nuclear and Radiological Leadership for Safety, launched in 2016. Cultivating a safety culture among staff, so that they can understand the importance of safety and the measures required to sustain it, is key in the nuclear industry. Establishing a strong safety culture is one the most fundamental management principles when using nuclear technology. It aims to strengthen the implementation of a systemic approach to safety, that is, the interaction between humans, technology and organizations within the national nuclear infrastructure. The importance of safety culture is one of the key lessons learned from the Fukushima Daiichi nuclear accident.
[en] The SSDL of Latvia was established from 2000 to 2001 with the financial and technical support of the IAEA. In 2002 it became a member of the IAEA/WHO SSDL network. The SSDL facilities are in Salaspils, just outside Riga, the capital of Latvia, on the territory of the former nuclear research reactor. The SSDL is part of a LEGMC Laboratory. Currently SSDL has 5 specialists (including head of Laboratory and quality manager) that are highly qualified and experienced professionals in the field of calibration and testing of ionizing radiation measuring and monitoring devices. There is Internal Quality Assurance system implemented in SSDL to guarantee required precision and accuracy of measurement results. The quality of services provided by the SSDL is ensured by a regular participation in international comparison measurements. The laboratory irradiators include the PANTAK PMC-1000 X ray irradiation unit (40 kV to 225 kV), as well as the gamma irradiators OB-2 (Co-60, 3.7GBq), OB-6 (Cs-137, 740 GBq) and panoramic gamma irradiator OB-34 (with four Cs-137 and three Co-60 sources) for the calibration and testing of protection and diagnostic radiology dosimeters and measuring devices.
[en] The accident at the Japanese nuclear power plant (NPP) Fukushima-1 in March 2011 showed that possibility of accidents with potentially serious radiation consequences could not be excluded with large-scale measures for improvement of safety level. For spent nuclear fuel storage facilities, one of such accidents may be the interruption of heat removal from spent nuclear fuel (SNF) due to the failure of the cooling system as a result of disruption of the power supply system with the failure of backup power sources or rapid full dehydration of the wet SNF storage as a result of the destruction of building structures and its depressurization. The decision to take preventive measures in advance to minimize exposure to personnel and the public is based on conservative estimates of possible radioactive discharges. To perform such assessments, the operating organizations carry out a calculated justification of the thermal and hydraulic characteristics of the SNF system in the accident scenarios with long-term blackout and a violation of heat removal. APROS is one of the software tools that are used in SEC NRS for calculating the thermal-hydraulic characteristics of systems in transient modes by solving the equations of heat and mass transfer in a steam-water mixture. For more detailed calculations of the structural elements of spent fuel assemblies (SFA) temperature, the ANSYS software is used, which implements the finite element method. The results obtained with the help of the above simulation tools are used by specialists of SEC NRS to assess the protective measures developed by operating organizations.
[en] An effective regulatory framework is essential to the success of a national nuclear power programme. The IAEA has developed the Milestones approach to help Member States embarking on nuclear power to understand and develop the necessary infrastructure requirements in a phased way. The regulatory framework is one of the 19 infrastructure issues that are described in the Milestones approach. The primary objective of this publication is to present the experiences of selected Member States that are in the process of building or expanding their regulatory framework for a nuclear power programme, including the challenges they faced. The publication also provides insights on IAEA safety requirements and guidance on establishing an effective regulatory framework with reference to the IAEA Safety Standards Series, the IAEA Nuclear Security Series, and IAEA Safeguards guidance publications. In addition, it demonstrates how these requirements fit into the overall development of a nuclear power programme through the IAEA Milestones approach.
[en] In Indian Pressurised Heavy Water Reactors (IPHWRs) calandria tubes are rolled with end shield tube sheet at either ends with the help of a sandwich type joints. These calandria tubes are generally kept unchanged during en-masse replacement of pressure tubes and end-fittings; probably due to unavailability of technology for replacement of calandria tube. But in recent past, one irradiated calandria tube has been replaced successfully from one of the 220 MWe IPHWRs, with a new one, with the help of Calandria Tube Rolled Joint Detachment (CTRJD) system, developed by Reactor Engineering Division of Bhabha Atomic Research Centre (BARC). This paper gives brief description of CTRJD system, methodology of calandria tube rolled joint detachment, shop floor trials of the technique for optimisation of operating parameters, qualification trials at full length mock up facility and deployment at reactor site. (author)
[en] The International Atomic Energy Agency (IAEA) and the Generation IV International Forum (GIF) have jointly committed to collaboration between their respective programmes, and to share information in selected areas of mutual interest. One of the key areas of emphasis in both the GIF and the IAEA programmes is the safety of liquid metal cooled fast reactors (LMFRs) including sodium cooled fast reactors (SFRs) and lead or lead-bismuth eutectic (LBE) cooled fast reactors (LFRs). A particularly important area of mutual interest is the harmonization of safety approaches, safety requirements, Safety Design Criteria (SDC), and Safety Design Guidelines (SDG) for the next-generation advanced LMFRs under development worldwide. This topic has gained increased importance in the aftermath of the accident that occurred in 2011 at the Fukushima Daiichi nuclear power plant, which drew renewed attention to nuclear safety and to the importance of an international safety framework for reactors currently in operation as well as for new designs.
[en] Conclusions: SDG-SSCs is a valuable exercise of consensus among designers at international level: useful for ''newcomers'' including TSOs and SAs; Few high ranked comments (16 over 134), but paragraph on containment design should be reframed; Many clarifications required - difficulty for TSO and SA to deal with design requirements; Objective to cover various concepts of SFRs makes SDG-SSCs sometimes difficult to apprehend (practical impact of requirements on safety).
[en] Summary - Introduction: ESFR SMART project is a four year project that began in September 2017.; It follows the Euratom CP ESFR project which was also a follow up of the European Fast Reactor (EFR) project.; Main purpose of the ESFR SMART project is to improve the reactor safety, and make a proposal for new safety options, based on both present and previous projects experience.; 1500MWe SFR pool type reactor with oxide fuel.; The deliverable giving the list of proposals for these new safety measures has been provided during the first year of the project with the drawings.; Several papers have been presented to explain these options in ICAPP 2018, ICONE 2018, ICAPP 2019, ICAPP 2020, Physor 2020; Reactivity control; Containment; Decay heat removal; Secondary loops and sodium fires detection; Conclusion: List of ESFR SMART simplifications - Dome (or polar table) suppression.; Safety vessel suppression, functions taken over by the reactor pit.; Primary sodium containment improvement with a massive metallic roof and other dispositions.; Natural convection cooling enhancement in the secondary side.; Optimized and simplified DHR dedicated circuits. (no DHX system in the primary vessel, no supplementary sodium circuits to manage).; Secondary loops with higher level of safety for sodium fires and sodium water reaction. Passive systems / Intrinsic safety - Passive control rods that stop the plant on physical parameters.; Low void effect in the core able to support severe transients (ULOF, etc.).; Passive decay heat removal by DHRS 2 and 1 (12 independent loops in natural convection) using only air, always available.; Thermal pumps totally passive to increase flow rate in natural convection and the decay heat removal systems capabilities,; Long delay before necessity of operator action, even in case of loss of water and loss of electricity supply.