The Exercise Evaluation software was developed by the SEC NRS to render the scientific and technical support to the Information and Analytical Centre of Rostechnadzor (Rostechnadzor IAC) — the Russian state safety regulation authority in the field of atomic energy.

The software is meant to be used by the Rostechnadzor and SEC NRS staff – members of the Rostechnadzor IAC working groups – to evaluate the preparedness and actions of the NPP operators in emergency drills and exercises. The software implements the recommendations of the ‘Guidelines for Evaluating the Effectiveness of Emergency Response Drills and Exercises Conducted by Nuclear Power Plant Operator’, approved by the Rostechnadzor Order No. 525 of November 7, 2013 (‘the Guidelines’). The criterion component of these Guidelines rests upon the requirements of relevant federal nuclear safety rules and regulations (‘Regulation on the Procedure for Declaring an Emergency, Prompt Reporting and Arrangement of Emergency Assistance to Nuclear Power Plants in Case of Hazardous Radiological Situations’ (NP-005-16), etc.). The final judgement philosophy implemented in the software is based on the recommendations given in the IAEA document Exercise 2005 “Preparation, Conduct and Evaluation of Exercises to Test Preparedness for a Nuclear or Radiological Emergency” (EPR-2005), and enables a quantified assessment of exercise / drill effectiveness.

The Exercise Evaluation software provides for emergency exercise / drill evaluation in 15 key emergency preparedness areas, including the completeness and correctness of operational information, the correctness of radiological consequences assessment for the accident, the size of protective measures, level according to the International Nuclear Event Scale (INES), timely notification of emergency response participants, and actions to evacuate personnel not involved in the accident management activities.

The introduction of the Exercise Evaluation software has brought a number of significant benefits for the Rostechnadzor IAC:

• minimized risk of incorrect INES classification of human performance events;
  easier application of the Guidelines;
• lower labour and time input to make the evaluation;
• automatic generation of a report containing the final judgement on exercise / drill effectiveness.

The software also features flexible display, storage and structuring of information on the results of emergency drills and exercises conducted by the NPP operators. Furthermore, the data base on exercise / drill results integrated into the software, allows comparing the nuclear plants in terms of diligent fulfilment of the requirements of federal nuclear safety rules and regulations while contributing to regulatory decision-making by Rostechnadzor.

Contact: Denis Mistryugov, SEC NRS, mistryugov@secnrs.ru
Source: ETSON

In recent years, there has been a growing interest across the world in the designs of small nuclear power plants (SNPP) with small reactors of various types. The Russian Federation has amassed ample experience in regulating the safety of marine nuclear power installations, in particular, in the course of the oversight and supervision over icebreakers operation.

The regulatory legal framework for regulating the safety of vessels and other watercraft with nuclear reactors, developed relying on this experience, takes into consideration the recommendations of the IAEA and IMO, the provisions of relevant international conventions and the requirements of national shipping regulations, and was used as a groundwork to put in operation the world’s first floating nuclear unit ‘Akademik Lomonosov’ which is in essence a floating nuclear power plant. However, the regulatory legal framework used for the floating SNPP is not applicable to the land-based small nuclear power plants because of a great number of requirements specific to stationary nuclear facilities.

Currently, the most promising Russian design of a land-based small nuclear power plant is that of a plant with the RITM-200 reactor system. The design is to be implemented in the framework of the comprehensive national programme ‘Development of Equipment, Technologies and Scientific Research in the Field of Atomic Energy in the Russian Federation for the Period up to 2024’ endorsed by the Russian President decree No. 270 of 16.04.2020.

The RITM-200 reactor plant has a unique integral configuration so that the core and the primary circuit components are placed directly inside the steam generation unit. The steam generation unit is quite compact and is fully assembled at factory, which ensures high-quality assembly and simplifies the unit transportation to the SNPP site.

Being a scientific and technical support organisation for Rostechnadzor, starting with 2019, SEC NRS has been providing a pre-licensing tracking of the land-based SNPP designs, including the RITM-200 facility. Thus, it looked at the safety philosophy of the land-based SNPP with the RITM-200 reactor, and made a preliminary assessment of the compliance of RITM-200 design solutions with the requirements of relevant nuclear regulations. The key objective of the study was to assess the adequacy of the existing regulatory requirements for regulation of nuclear, radiological and industrial safety of SNPP with RITM-200 reactor facility. The analysis has confirmed that the documents regulating the safety of large NPPs are suitable for regulating the RITM-200 SNPP safety. The analysis also identified the RITM-200 SNPP design features that may require amendments in the current federal nuclear safety rules and regulations and the improvement of relevant standards. At the same time, the designers were advised of the areas requiring further demonstration of the RITM-200 SNPP safety or changes in the design documentation.

Contact: Denis Mistryugov, SEC NRS, mistryugov@secnrs.ru
Source: ETSON

Similar to the situation with other human activities, the coronavirus pandemic has put to test the regulator ability to adapt and retain effective performance under extreme conditions. The emergency procedures are normally developed to respond to human-induced disasters, and to the nuclear facility accidents initiated by natural or human-made events. In the face of the coronavirus, these procedures require a scrupulous adjustment to accommodate the new challenges posed by the pandemic.

The Regulator is challenged primarily to:

• protect the life and health of its own staff (both in headquarters and regional offices), and the staff of the operating organisations ensuring safe operation of the nuclear facilities under the coronavirus pandemic;
• arrange effective remote work, including conduct of remote inspections to ensure continued oversight of the nuclear and radiological safety at supervised facilities.

Rostechnadzor developed and has been implementing the following measures to meet these challenges:

• the number of the Rostechnadzor on-the-job staff has been reduced to as low as practical. All these people have been provided with personal protection items, their social distancing has been secured, and they have been regularly tested for coronavirus;
• remote work has been arranged for the rest of the Rostechnadzor staff through state-of-the-art digital communication tools, to ensure timely receipt of necessary information and provide access to data bases as needed for their efficient work.

Apart from that, the regional offices of Rostechnadzor were instructed to:

• ensure appropriate scope of inspections and other activities to enable verification of the nuclear plant safety by the inspectors authorized to perform continuous state supervision;
• introduce supervision over the implementation of the additional measures developed by Rosenergoatom to ensure the safe operation of the nuclear power plants under the high alert regime imposed on the territory of the Russian Federation.

Also, the Rostechnadzor guided the operating organisations to:

• ensure safe operation of the nuclear power plants under current restrictions;
• develop additional actions to ensure the operational safety of the nuclear power plants, in particular:
  • increase the number and thoroughness of the process system / equipment rounds;
    deliver extra briefings for personnel to underline the work specifics under the high alert regime;
  • minimise scheduled lineups and maintenance of equipment, except for emergencies;
    ensure the safety of the operating and duty personnel at the nuclear power plants in compliance with relevant public health requirements;
  • develop measures to warrant availability of at least the minimum number of operating personnel justified in the design documentation, should part of the staff be taken to hospitals.

Contact: Denis Mistryugov, SEC NRS, mistryugov@secnrs.ru

Source: ETSON

A leak suspected to have been caused by high cycle thermal fatigue due to fluctuating axial motion of a swirl penetration front end has been detected in early 2018 in a non-isolable Reactor Coolant System branch at the Doel 1 Nuclear Power Plant (Belgium).

As a part of the standard design of the Westinghouse 2-loop PWRs, the Doel 1 and Doel 2 units are equipped with ECC systems that inject water into the upper plenum of the reactor pressure vessel in the event of a LOCA, as opposed to the cold-leg injection that is also present. Each of both reactors has two 4-inch UPI lines made of Type 316 stainless steel and further referred to as UPI-A and UPI-B lines. Figure 1 illustrates the layout of the UPI-A lines. When compared to the UPI-A lines, the UPI-B lines are shorter and have a slightly different slope.

The UPI injection ports are at the same elevation as the RPV inlet and outlet nozzles. In the initial design of the Doel 1 and Doel 2 units, the UPI lines provided ECC water in the lower plenum by injection in the downcomer but in 1992 ther injection ports were re-configured to the UPI design by plugging the opening towards the downcomer and drilling a hole into the core barrel. Figures 2 and 3 illustrate the design of the injection ports in both configurations.

On the 23rd of April 2018, a primary leak was detected at Doel 1 while the unit was in full power operation. Consequently, the unit was brought to cold shutdown conditions. The leak was found to be located on the UPI-A line inside the reactor cavity in the pipe section downstream of the check valve, i.e., in a non-isolable RCS branch. Extensive NDE of the damaged UPI-A line including mechanized VT, ET, UT, and UT-TOFD from the inside of the pipe has evidenced several cracklike indications in the bottom part of the straight part of the pipe upstream of the weld between the straight section and the elbow close to the leak location, as well as a circumferential indication in a neighboring weld. As a result of those findings, the same NDE was performed on the UPI-B line of Doel 1 and on both UPI lines of Doel 2. These inspections highlighted similar degradation on the UPI-A line of Doel 2 at the same location. By contrast, no reportable degradation was observed on the UPI-B lines.

The affected section the UPI-A lines was removed for further laboratory investigations aiming at investigating the possible mechanisms responsible for the formation of the cracks. The outcomes of the investigations allowed the plant Licensee (Engie) to conclude that the mechanism responsible of the cracking phenomenon was fatigue. More specifically, different features heavily suggested that the mechanism is a low stress, high-cycle fatigue mechanism, possibly environmentally (primary water) assisted. Root cause analysis performed by the Licensee concluded that this phenomenon would have likely been related to the occurrence of cyclic thermal stresses in the wall thickness of the piping. A detailed understanding of how the damage could develop, implying in particular a quantitative description of all involved phenomena, could however not be provided.

The removed sections of the UPI-A lines were replaced and a specific Safety Case has been provided by the Licensee for each unit, as a justification for their restart. Despite an incomplete understanding of the degradation phenomenon, the Licensee justified the allowance for restart on the basis of different actions among which of the replacement of the damaged UPI-lines, the planning of NDE at regular intervals as well the continuous monitoring of those lines and in particular the temperature monitoring of the external wall temperature at specific locations by means of the FAMOSi Data acquisition system. In answer to Bel V’s concern, the Safety Cases were based on the application of the Defence in Depth principle, by showing how the actions defined in the Safety Cases are related to the first three Levels. After evaluation, Bel V concluded that the Doel 1 and Doel 2 plant might be operated safely during a time period of one fuel cycle, which is expected to be the time necessary to identify and quantify the thermal cyclic loadings suspected to be the mechanism responsible for the detected cracking. Authorization for restart was granted by the Belgian Federal Agency for Nuclear Control and both plants resumed operation early in 2019.

During the 2019 fuel cycle, two major achievements should be highlighted. First, the temperature measurements recorded by the FAMOSi system during four different plant conditions, i.e. cold shutdown, heat-up, hot stand-by, power ramp-up and full power were made available and analyzed by the Licensee by means of a homemade computer code allowing to estimate the temperature evolution of the fluid inside the pipe from the recorded temperatures. Although the FAMOSi system has experienced reliability issues, some conclusions were able to be drawn from the temperature records. Temperature fluctuations were observed on all UPI lines as soon as at least one reactor coolant pump is running, whatever the plant condition. However, those temperature fluctuations are not significant , which led the Licensee to conclude that the thermal cycling which was suspected to be at the origin of the observed cracks is rather a consequence of cycling flow structures inside the lines. Secondly, the results of Computational Fluid Dynamics (CFD) calculations performed by the Licensee have been issued. Those calculations point out that cycling flow structures could develop in UPI-A lines in the presence of a gap between the reactor vessel UPI nozzle and the core barrel. The presence of this gap, which has been confirmed during the visual examinations carried out on the inside of the UPI lines, is a side effect of the reconfiguration from downcomer injection to upper plenum injection. The pressure difference existing between the downcomer and the upper plenum generates a flow through the gap, which then induces a swirling structure extending in the UPI-A line up to the degraded region if adequate non-symmetric flow ingress through the gap is assumed. Beyond that swirl penetration zone, a natural circulation loop with stratification is present. Between those thermal fields, a transition zone occurs with potentially high temperature difference. The CFD calculations also confirm that, if slightly varying gap boundary conditions are assumed, the extension of the swirl region moves back and forth, which induces large temperature fluctuations in the degraded region. With regard do that, it should also be pointed out that the assumed mechanism also explains why the UPI-B line is not or less affected than the UPI-A line. The better thermal insulation on this former line indeed induces a smaller thermal axial gradient along the line, especially in the natural circulation loop zone. This explains why smaller thermal fluctuations will occur when the extension of the swirl zone moves back and forth.

However, convincing experimental demonstration could not be achieved due to lack of temperature measurement sections at some critical locations but also to a lesser extent, due to the reliability issues encountered by the FAMOSi system. In order to hopefully conclude about the identification of the mechanism responsible for the damage phenomenon and the explanation of the occurrence of the damage phenomenon under the assumed mechanism, the Licensee decided to remedy the malfunctions of the FAMOSi system and to install additional measurement sections during the next refuelling outage of the units.

During the 2019-2020 outage, several non-destructive inspections have been carried out on the UPI lines of Doel 1 and Doel 2, according to the programme agreed with the Safety Authorities. Depending on the part of the line, VT, UT, LP and/or RX techniques have been used. No reportable indication has been detected.

In March 2020 an Addendum Safety File was issued by the Licensee, which summarizes the results from the most recent investigations, reassesses the Root Cause Analysis and justifies the safe operation of both plants Doel 1 and Doel 2. The request of the Licensee to allow the restart of the Doel 1 and Doel 2 unit was accepted by the Federal Agency for Nuclear Control, with due consideration of the Addendum Safety File and the inservice inspection program of the UPI lines.

A first assessment of the temperature monitoring results is expected by the end of September 2020.

Contact: Guy Roussel, Bel V, guy.roussel@belv.be

Russian federal nuclear safety rules and regulations ‘Basic Safety Provisions for Nuclear Power Plants’ (NP-001-15) establish a requirement to complement the safety analysis for a nuclear power plant with estimation of errors and uncertainties in the obtained results. Similar requirements can be found in regulatory documents of nuclear safety authorities in most of the countries with well-developed nuclear programme, as well as in the IAEA document GSR Part 4 ‘Safety Assessment for Facilities and Activities’.

Over the past three decades, the Russian Federation, the same as many other countries, has accumulated a great deal of experience in the application of uncertainty estimation techniques for the results of the calculations performed to demonstrate various aspects of the nuclear plant safety. However, these methods are usually discussed in scientific papers published in journals, or in presentations made at conferences, seminars and meetings, while their actual application in respect of NPP safety documentation is very limited, despite the above requirements enshrined in the nuclear safety authority regulations. This may be caused by the lack of national regulatory documents that would guide the safety documentation authors on how they should fulfil the requirement to estimate errors and uncertainties in the results of computational safety analyses.

With this in mind, the SEC NRS had developed a Nuclear Safety Guide ‘Guidance on Estimation of Errors and Uncertainties in the Safety Analysis for Nuclear Power Plants’ (RB-166-20) that was approved by the Federal Environmental, Industrial and Nuclear Supervision Service in its the order No. 288 of July 30, 2020.

The RB-166-20 Safety Guide contains the recommendations on estimation of errors and uncertainties in the results of a deterministic safety analysis performed using the software tools to demonstrate safety of a nuclear power plant. Thus, the Federal Environmental, Industrial and Nuclear Supervision Service has become a pioneer in issuing uncertainty estimation recommendations.

Before starting the RB-166-20 development, the authors made an overview of the world’s best practices in the assessment of errors and uncertainties in the results of the computational safety analyses for nuclear power plants. The study was carried out by SEC NRS together with the staff of the Institute for Safe Development of Nuclear Power with Russian Academy of Sciences, the National Nuclear Research University ‘MIFI’, and the A.P. Aleksandrov Science and Research Technological Institute which is part of the State Atomic Energy Corporation ‘Rosatom’. Noteworthy is the significant contribution made in the RB-166-20 drafting by the staff of the OKB GIDROPRESS, NIKIET, Afrikantov OKBM, Research Centre ‘Kurchatov Institute’, and VNIIAES.

The RB-166-20 consists of two main sections. The first section provides guidance on accounting for uncertainties in the design safety analysis for a nuclear power plant. The second section gives recommendations for the use of uncertainty evaluation methods. The Annexes to the RB-166-20 contain a block diagram showing the uncertainty estimation process for the results of the computational analysis of the design-basis accidents at a nuclear power plant, and a description of the approach to the assessment of errors and uncertainties in the results of beyond-design-basis accident calculations for nuclear power plants, including severe accidents.

The Safety Guide is meant to be used by the staff of operating organisations, Rostechnadzor divisions, and other institutions conducting activities in the field of nuclear energy.

Contact: Denis Mistryugov, SEC NRS, mistryugov@secnrs.ru

The reactor pressure vessel (RPV) is the most critical component of light water reactors with regard to plant safety and lifetime. The Nickel-base Alloy 182 is widely used as a weld filler and attachment pad metal to join the low-alloy steel RPV and RPV nozzles to both wrought nickel-base alloys and austenitic stainless steel components (piping, reactor internals) in light water reactors by manual shielded metal arc welding. The recent stress corrosion cracking (SCC) in Alloy 182 in RPV nozzles, reactor internal attachment or penetrations dissimilar metal welds in boiling water reactors (BWRs) represent a serious potential safety concern (Figure 1). [1]

The Alloy 182 weld metal is very prone to SCC and cracks tend to grow towards the RPV due to weld microstructure and residual stress profiles in these welds. Such SCC cracks thus might grow into the adjacent RPV, although the SCC was confined to the weld metal and none of the SCC cracks significantly penetrated the adjacent RPV base material in field cracking incidents so far. This is consistent with the very high SCC resistance of the RPV steel. Even after post-weld heat treatment, these highly constrained welds reveal high residual stresses resulting in a high mechanical driving force for SCC cracks that are reaching the fusion boundary region between the weld metal and RPV steel. There is thus a concern that such cracks could grow into the RPV in oxidizing BWR normal water chemistry (NWC) environment, in particular in high-sulphur RPV steels (older plants) or in case of chloride contamination of the coolant or weld repairs [1-3]. It is thus crucial to know, under which conditions these cracks could grow into the RPV. Furthermore, validated SCC crack growth disposition curves for the RPV steels are needed for flaw tolerance and integrity assessments and verification of inspection intervals of periodic in-service inspection and to demonstrate sufficient safety margins. [1]

Within several projects that were funded by the Swiss Nuclear Safety Inspectorate (ENSI), LNM has investigated these aspects by very challenging, time-consuming and costly experiments [1-4] and developed new SCC disposition curves in collaboration with the Electric Power Research Institute (EPRI) in the USA [5]. In high-purity water (< 1 ppb of chloride, 1 ppb = 1 μg/kg), SCC crack arrest was observed arrest at the fusion boundary up to high stress intensity factors (Figure 2). On the other hand, small chloride contamination of > 3 ppb caused fast SCC into the RPV down to low stress intensity factors values in BWR/NWC environment! It is thus essential to maintain a high-purity of the reactor water, and the Action Level 1 limit of the EPRI BWR/NWC water chemistry guidelines was reduced to 3 ppb based on these investigations in 2016.

In BWRs with reducing hydrogen water chemistry (HWC) (as it is applied in the Swiss BWR KKL and KKM) with hydrogen injection into the feedwater, a much higher chloride tolerance was observed (Figure 3), where no SCC occurred in the RPV steel up to high chloride concentrations of 500 ppb.

Based on our results and suggestions, new SCC dispositions crack growth curves for RPV steels in BWR/NWC and HWC environment (Figure 4) and during chloride transients (Figure 5) were developed in collaboration with EPRI [5] and implemented in the Code Case N-896 in the ASME Boiler and Pressure Vessel Code in 2019 that were also accepted by the US regulator NRC. These curves are used for flaw tolerance evaluations and integrity assessments of the RPV. With these curves, sufficient safety margins for the RPV could be demonstrated with the current inspection intervals of the periodic in-service inspection for high-purity water. In case of HWC, margins are extremely high, even in case of moderate chloride contamination. It took almost 10 years from the first ideas (PSI publication in 2008) [4] to the acceptance of the Code Case in 2019. The Code Case is an important milestone and contribution to the safe long-term operation of BWRs.

The principal investigator H.P. Seifert has recently received the Coriou Award 2019 of the European Federation of Corrosion (EFC) in recognition of his significant contributions to nuclear corrosion and SCC in the last two decades (Figure 6).

Original publications

1. H.P. Seifert, S. Ritter, H.J. Leber, S. Roychowdhury, Stress Corrosion Cracking Behavior in the Transition Region of Alloy 182/Low-Alloy Reactor Pressure Vessel Steel Dissimilar Metal Weld Joints in Light Water Reactor Environments, Corrosion 71 (2015) 433-454.
2. H.P. Seifert, S. Ritter, The influence of ppb levels of chloride impurities on the stress corrosion crack growth behaviour of low-alloy steels under simulated boiling water reactor conditions, Corrosion Science 108 (2016) 134–147.
3. H.P. Seifert, S. Ritter, The influence of ppb levels of chloride impurities on the stain-induced corrosion cracking and corrosion fatigue crack growth behaviour of low-alloy steels under simulated boiling water reactor conditions, Corrosion Science 108 (2016) 148–159.
4. H.P. Seifert, S. Ritter, Stress Corrosion Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions, Journal of Nuclear Materials 372 (2008) 114 – 131.
5. S. Ranganath, R.G. Carter, R. Pathania, S. Ritter, H.-P. Seifert, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment, in: Proceedings of the ASME 2018 Pressure Vessels and Piping Conference PVP2018 July 15-20, 2018, Prague, Czech Republic, Paper PVP2018-84257.

Contact: Hans-Peter Seifert, Laboratory for Nuclear Materials (LNM), Paul Scherrer Institut, hans-peter.seifert@psi.ch
Source: ETSON

One of the key duties of a nuclear safety authority is to ascertain the nuclear facility safety. Computational modeling of credible operational transients and accidents that may occur at a nuclear facility is imperative for demonstrating its nuclear and radiological safety. Software developers conduct a broad spectrum of computational and experimental research to demonstrate the computer code eligibility for modeling a certain facility – this procedure is usually referred to as ‘verification and validation’ of software. The outcome of the verification and validation process is reviewed by experts to confirm that the V&V has been performed as appropriate.

In the Russian Federation, the expert review of the software is a mandatory pre-licensing process that provides for an in-depth and comprehensive assessment of the calculation techniques used in the safety analysis. Furthermore, the review is conducted with the involvement of the Rostechnadzor’s Expert Board whose members are top computational modeling experts from SEC NRS, Rosatom organisations, national research centres, universities and Russian Academy of Science institutions.

The coronavirus pandemic made it necessary to give another thought to the communication formats used in the software review process, to expedite introduction of front-edge digital technologies. The need to arrange remote access activities stimulated the rapid transfer of the main code review sub-processes to the SEC NRS portal https://soft.secnrs.ru/ which uses Russian Software platform «1С-Bitrix24». The following was done:

1. A 24-hour interactive access was arranged for the reviewers to get updates on the code review progress.
2. All developments in the review are posted in the ‘Live news line’, and are forwarded to each external user as an e-mail notification. The flexible setting of rights provides for the user access to the required information.
3. The reviewers work in secured groups provided with interactive communication arrangements (chats or videoconferences, group or ‘one-on-one’). All required documentation is kept in a centralised location. The reviewers upload digital documents at the portal. The implemented planning tools include the Gantt chart, Kanban and others. The software is tested on-line (14 computer codes have been tested so far).
4. The certificates bank is fitted with an upgraded ‘Quick Search’ tool and a user-friendly sorting interface.

By today, we have accomplished secured interactive communication of experts. As a next step, we will move to the second, more complicated, stage of digitalisation– to the introduction of automatic expert review reports and analysis of the results obtained.

Contact: Denis Mistryugov, SEC NRS, mistryugov@secnrs.ru
Source: ETSON

MYRRHA (Multipurpose hYbrid Research Reactor for High-tec Applications) is a unique highly innovative prototype of a nuclear reactor driven by a high power linear accelerator. It will be designed to operate in both critical (no accelerator needed) and subcritical mode. MYRRHA has a fast spectrum core cooled by Lead Bismuth Eutectic (LBE) (See Figures 1 and 2).

The primary purpose of MYRRHA is to proof the concept of transmutation of minor actinides. If successful, this process could be implemented at industrial scale allowing for a reduction on the processing and storage time of nuclear waste. Other applications are also envisaged for MYRRHA, such as, the production of new radioisotopes for the treatment of diseases, the testing of materials for nuclear fusion and fission reactors, the provision of operating experience and data for the development of advanced nuclear reactors and the conduct of pioneering research.

The licensing and construction of such an innovative nuclear device is a long-term endeavor. Currently the design is undergoing a pre-licensing phase. Pre-licensing is a process that Belgian prospective licensees can apply for in any project, with the aim of identifying at an early stage barriers to the licensability of the design. The objectives of the pre-licensing process are:

• The identification by the nuclear safety authorities of the necessary improvements to the regulatory framework in relation with the future licensing application.
• The elaboration by the nuclear safety authorities of the safety and security objectives expected from the future facility.
• The review by the nuclear safety authorities of the feasibility and the licensability of the project.
• The description, by the designer, of the safety and security options that apply to the future facility in view of satisfying the minimum objectives expected by the nuclear safety authorities.
• The elaboration, by the designer, of answers and justifications to specific points of attention identified early in the process.

MYRRHA is a very innovative design where the designer is currently undertaking significant R&D programs in order to understand the technological and scientific challenges posed by the innovative design decisions made. Bel V will have to assess the solutions applied to those challenges in terms of safety for the people and the environment, which poses a different challenge. Bel V’s highly qualified engineers and scientists get ready for this challenge by:

• Participating in international forums, such as the OECD/NEA Working Group on the Safety of Advanced Reactors where matters related to the licensing of advanced technology are discussed by regulators from different countries.
• The elaboration of internal topical reports on specific innovative areas with the aim of identifying current knowledge gaps, and the way in which they can be filled by training, R&D or working with experienced partners.
• Bilateral contacts with other regulators to discuss lessons learnt in the application of their respective (pre-)licensing processes to innovative designs.
• Developping experience & competence in computational tools and methods for metal cooled fast reactors.

Ultimately, MYRRHA offers the possibility of training a next generation of engineers and scientists in the nuclear field, together with experienced professionals, by participating from the onset of this project in understanding and working out solutions to the many challenges of licensing advanced nuclear technology.

Contact: Gustavo Rubio Anton, Bel V, gustavo.rubioanton@Belv.be.
Source: ETSON.

Группа экспертов МАГАТЭ завершила 11-дневную миссию по рассмотрению мер по обеспечению готовности и реагированию в Канаде в случае ядерных и радиологических аварийных ситуаций. Обзор готовности к чрезвычайным ситуациям (Emergency Preparedness Review – EPREV) является одним из экспертных обзоров, предлагаемых МАГАТЭ для усиления ядерной безопасности в государствах-членах.

Обзор EPREV сосредоточен на мерах и возможностях для подготовки и реагирования на ядерные и радиологические аварийные ситуации. Обзор EPREV основан на стандартах безопасности МАГАТЭ в области ядерной и радиологической аварийной готовности. Миссия в Канаде была сосредоточена на обеспечении готовности к чрезвычайным ситуациям в результате событий на АЭС. В Канаде эксплуатируется 19 реакторов на четырех площадках, которые обеспечивают около 15% электроэнергии.

Группа проверки обзора EPREV, состоящая из 11 человек, определила несколько сильных сторон в ходе миссии, включая хорошо разработанную и зрелую систему реагирования на чрезвычайные ситуации и обеспечения готовности на всех уровнях управления. Правительство разработало оптимизированный подход для своевременной обработки требований об ответственности, связанных с ядерными или радиологическими аварийными ситуациями.

В рекомендациях для дальнейшего рассмотрения указано, что правительство должно включить обоснование и оптимизацию в стратегию защиты и разработать детальную стратегию мониторинга для оптимизации использования возможностей и ресурсов мониторинга. Правительство должно также разработать подробные меры для прекращения ядерной аварийной ситуации.

Подготовка канадским правительством к этому обзору была четкой, целенаправленной и эффективной. Выводы этой миссии помогут Канаде в дальнейшем укреплении ее системы аварийной готовности и реагирования.

Обзор EPREV был проведен по запросу правительства Канады, которое, по сообщению МАГАТЭ, намерено принять план действий для учета полученных результатов. Правительство планирует опубликовать отчет после его завершения в начале 2020 г. Канада проведет последующую миссию обзора EPREV примерно через два-четыре года. Канада была первой страной с крупной ядерно-энергетической программой, которая провела обзор EPREV.

Источник: World Nuclear News, 14.6.2019