12.09.2022 16:20 Project JEK2

Project JEK2 – 101

Status of the Project for the Construction of a New Nuclear Power Plant JEK2

Bruno Glaser

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia


The electricity generation sector is very important for Slovenia and the EU, so it is necessary to strive for a stable, efficient, and sustainable energy supply. At the end of 2020, the EU proposed raising the target for reducing greenhouse gas emissions by 2030 to at least 55% compared to 1990, which was also confirmed by the European Commission, and the EU plans to phase out fossil fuels by 2050.
The national strategy in Slovenia includes, in addition to efficient energy use, promoting the use of renewable sources, promoting cogeneration, and preventing climate change, guidelines on projects to achieve sustainable development of Slovenia such as extending the operational lifetime of NEK and building a new Krško Nuclear Power Plant (JEK2). Given the excellent experience in the use of nuclear energy, the choice and continuation of the nuclear option is a natural choice.
With the adoption of the National Energy and Climate Plan (NEPN) and the Climate Strategy in Slovenia, GEN gained a strategic basis for the continuation of the JEK2 project. The key steps that have taken place in the recent period are obtaining an Energy permit in July 2021, preparing and submitting documentation for the initiative to start the spatial planning to the Ministry for Infrastructure as the initiator and handing over the application of the initiator to the Ministry for environment at the end of March 2022. Several site investigation studies are in progress like Seismic and geology, Analysis of site selection variants, Analysis of the operation of cooling towers and their impact on the environment, in preparation for the environmental report within the Environmental Impact Assessment. Other major activities are related to the preparation of the requirements for suppliers to start negotiations (Request for Vendor Information), projection of staffing requirements and preparation for the establishment of JEK2 organization, participation in European Utility Requirement group (EUR) design assessments and SNSA licensing (nuclear licensing of JEK2 project).

12.09.2022 16:40 Project JEK2

Project JEK2 – 102

The Westinghouse AP1000® Plant – Proven, Advanced Generation III+ Technology

Elias Gedeon

Westinghouse Electric Company UK Ltd., Springfields, Salwick, Preston PR4 0XJ , United Kingdom


The AP1000 plant is the most advanced yet proven nuclear power plant technology available, with four units breaking operational performance records for plant availability, flexibility, and short refuelling outage durations. Two more units are expected to load fuel in 2022/2023 and a further four units have now received construction approval. An AP1000 plant project at Krško would be at least number 11 in the series. The experience of the first and second wave of projects would be incorporated at Krško to ensure the AP1000 plant is effectively delivered to become the most proven, advanced generation III+ technology option for the Slovenian new build program and beyond. In the 1990s, utilities and regulators demanded an improved Generation III+ class of reactor, achieving increased levels of safety, constructability, and operability, while reversing the trend for greater complexity in reactor design. The AP1000 pressurized water reactor sets a new standard for nuclear power plant simplicity and safety using all passive safety systems, requiring no AC power or operator actions to maintain plant cooling for at least 72 hours after a postulated accident event. The AP1000 plant also utilizes advanced, modular construction methods, further optimized for the efficient delivery in Slovenia and all future projects.

12.09.2022 17:00 Project JEK2

Project JEK2 – 103

EDF EPR 1200: the European solution to support Slovenia’s energy transition

Marie Agnes Berche

Electricité de France, Generation and Transmision, Site Cap Ampere, 93207 Saint Denis Cedex, France


Based on our 2000+ years operating experience in PWRs, EDF has developed the EPR, the most mature GEN3+ in the world. With its European partners, EDF is now building a fleet of EPRs across the continent in a context where strategic countries such as Slovenia are considering new nuclear capacities.

12.09.2022 17:20 Project JEK2

Project JEK2 – 104

APR1000, Best Solution for Slovenian Nuclear New Build Project

Ji-Yong Oh, Won-Seok Yang, Keunho Lee

KHNP – Korean Hydro &Nuclear Power co. ltd, Ulchin Nuclear Power Site Unit 6, #84-4, Bugu-Ri, Buk-Myeon, Ulchin-Gun, 161-101, South Korea


KHNP is currently operating 24 different types of reactor at 5 sites.Ever since Kori unit 1 was introduced in 1971, Korea has continued to construct nuclear power plants and KHNP can cover the entire cycle of an NPP project from project planning, design and engineering, to manufacturing, construction, and commissioning. We have developed the so-called GEN?+ reactors such as APR+, EU-APR and APR1000 through technology advancement and independence. The main components of APR1000 are designed to have 60 years of life time. Major safety systems are designed with full 4 trains of 100% mitigation capacity considering redundancy and OLM. In addition, APR1000 introduced passive auxiliary feedwater system that provides significant capability for the station black out. Further information on APR 1000 will be delivered at the NENE 2022 Conference.

12.09.2022 18:00 Education and training

Education and training – 202

ENEN2plus, Building European Nuclear Competence through continuous Advanced and Structured Education and Training Actions

Gabriel Lazaro Pavel1, Roberta Cirillo1, Csilla Pesznyák2

1European Nuclear Education Network, Rue d’Egmont 11, 1000 Brussels, Belgium

2Budapest University of Technology and Economics, Mûegyetem rkp 3-9, Budapest 1111, Hungary


Education and Training (E&T) actions in the nuclear field in Europe need persistent efforts to be adequately promoted, aiming to maintain and further develop the high level of expertise reached so far. This, also considering the limited attractiveness of nuclear careers for young generations that is currently experienced at Universities and recruiting for jobs. The European Nuclear Education Network (ENEN), since its establishment in 2003, has had “the main purpose of the preservation and the further development of expertise in the nuclear fields by higher education and training”.
For almost twenty years, ENEN has deployed efforts supporting E&T in nuclear sectors, with the continuing contributions to EU funded projects, organising the most important actors in the E&T and cooperating with technological platforms and industrial bodies. Projects ANNETTE and ENENPlus (both led by ENEN) financed student mobility at an unprecedented level in the field of nuclear fission, moving more than 500 learners for a total experience exceeding 43 person-years, and in reaching out at the levels of secondary school, BSc, MSc and PhD students.
This created an improved attractiveness for nuclear careers, with a clear benefit for Europe, confirming that Europe is a region in which nuclear studies can be developed at a high level also exploiting the excellence achieved by the nuclear industry and the medical applications.
The ENEN2plus project makes use of the experience gained in these recent endeavours, to continue the actions of ENEN in favour of nuclear E&T, in cooperation with a wide Consortium of qualified institutions.

12.09.2022 18:20 Education and training

Education and training – 203

Nuclear education and knowledge management activities at the NEA

Antonella Di Trapani1, Tatiana Ivanova2, Michael Fleming2, Alice Dufresne2, Oliver Buss2

1OECD Nuclear Energy Agency, Le Seine St. Germain; 12, Boulevard des Iles, 92130 ISSY-LES-MOULINEAUX, France

2NEA Data Bank OECD, 12 bd des Iles, F- 92130, Issy-Les-Moulineaux, France


Oliver Buss, Antonella di Trapani, Alice Dufresne, Michael Fleming and Tatiana Ivanova
OECD Nuclear Energy Agency (NEA), Paris, France
Corresponding author:

The NEA has been active in addressing issues associated with education, training and knowledge management needs for many years.
The 2000 OECD/NEA report “Nuclear Education and Training: Cause for Concern?” flagged the magnitude and urgency of the issue to governments. Although some actions had been taken and improvement noticed, strains in the human resources capacity remain strong.

The Division of Nuclear Science and Education is responsible for the developing and implementing the activities in the area of nuclear education and knowledge management at the NEA which will be presented in this paper.

The Nuclear Education, Skills and Training (NEST) Framework was launched in 2019 with the aim to develop skills and competences for the next generation of subject-matter experts through hands-on activities carried out in challenging multi-national projects. NEST Fellows are paired with NEST Mentors who are experts in the field, in order to facilitate the knowledge transfer to the young generation. Six projects in diverse areas of nuclear energy are currently running with the aim to train over 200 NEST Fellows in the next three years.

The Global Forum on Nuclear Education, Science, Technology and Policy is a new NEA initiative aimed at creating an inclusive network of experts from academia, who are responsible for nurturing the next generation of nuclear leaders. It will enable the exchange of ideas and facilitate dialogue on some of the most pressing issues the nuclear sector is facing today, such as how to improve some of the nuclear education and knowledge management issues by providing policy advice and best practices.
The Data Bank is responsible for the collection, preservation and dissemination of nuclear data, developing tools assisting in the validation as well as providing a benchmarking of the data. It organises training courses on the computer codes widely distributed and used in the Data Bank participating countries.
Around 10 training courses are organised every year, each one attended by 10 to 20 participants. These training courses help Data Bank participating countries maintaining a high-quality professional education and addressing knowledge management issues.

Other activities are carried out within the different divisions of the NEA in the form of international schools (nuclear law, radioprotection), or workshops (radioactive waste management, human aspects, nuclear science) and these will also be briefly presented.

12.09.2022 18:40 Education and training

Education and training – 205

Excellence in operator fundamentals at NEK

Matjaž Žvar

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia


Operator fundamentals are defined as the essential knowledge, skills, behaviours and practices that individuals and operating crews need to apply to operate the plant effectively. The fundamentals that all operators should demonstrate are as follows: understanding of plant design and system interrelationships, monitoring plant indications and conditions closely, controlling plant evolutions precisely, approaching conservative to plant operations and having an effective teamwork.

The operator fundamentals are essential to prevent production losses or significant events to occur in nuclear power plants. The continuous usage of human error reduction tools and improvements in different processes are causing reliable and steady state operation of the power plants, preventing operators to gain experience and by so growing an excellent path to their knowledge loss, if training is not reinforced. This is a worldwide known degradation, also recognized by the World Association of Nuclear Operators – WANO, after events that have had happened worldwide in the industry.

At NEK we are coping in different ways of operations fundamentals monitoring to achieve excellency. We are gaining information from the as-found and from the as left simulator scenarios where the crews perform their job under different plant conditions – normal operation, during transients or emergency operation, under the supervision of their superiors and instructors. This data is then analysed and feedback is given to the operating crews and their superiors in different ways, also using performance indicators on operator competencies. But the main driver to achieve excellency in operator fundamentals is the active involvement of all participants – the operating crews, the operations management and the training department.

The paper describes the comprehensive processes established by the training department at NEK to prevent events related to operator fundamentals.

13.09.2022 8:30 Research reactors and radiation measurement

Research reactors and radiation measurement – 300

The role of research reactors to enhance NPP fleet performance and safety

Patrick Blaise

French Atomic and Alternative Energies Commission (CEA), Saclay, France


The presentation focuses on the fundamental role of critical facilities in the enhancement of safety and performance of the current and forthcoming NPP fleet, through the experimental validation of both numerical schemes and nuclear data. The main approaches used to design and conduct experiments will be covered. The talk will be illustrated by several experimental programs performed in the French zero power reactors (ZPR) and their feedbacks for the global enhancement of the fleet. The presentation will resume what are the expected needs for the next generation of advanced reactors, and, given the current ZPR landscape worldwide, what are the potentialities in a broader international collaboration around existing facilities. Some highlights on a shared new ZPR, versatile, to consolidate multipurpose and analytical experiments will also be introduced.

13.09.2022 08:30 Research reactors and radiation measurement

Research reactors and radiation measurements – 301

Ex-core neutron measurements with a SiC-based diode and a sCVD Diamond based detector in JSI TRIGA Mark II research reactor

Valentin Valero1, Vladimir Raduloviæ2, Luka Snoj2, Laurent Ottaviani1, Abdallah Lyoussi3, Christophe Destouches4, Adrien Volte1, Michel Carette1, Christelle Reynard-Carette1

1Aix-Marseille Université, 58, bd Charles Livon, 13284 Marseille Cedex 07, France

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

3CEA-Cadarache DEN/DER/SPEX/LDCI, Bâtiment 238 – Piece 10, F13108 Saint-Paul-lez-Durance, France

4Commissariat a l’Energie Atomique – Centre d’Etudes de Cadarache DRN/DER, Bat 238, 13108 St Paul Lez Durance Cédex, France


Since many years, in the field of nuclear instrumentation, there is a need to develop new sensors for key-parameter measurements. These sensors have to be adapted for experiments carried out in nuclear fission or fusion facilities with restricted spaces, high power densities, high absorbed dose rates, high neutron and gamma fluxes. The main characteristics expected for these detectors are a small size, a fast response, a high sensitivity, a stability versus radiations, a strong discrimination between neutrons and gammas, and a high energy resolution. In fact, in order to meet these requirements, in particular for neutron measurements, a new technology based on wide-bandgap semiconductors is under development. Among wide-bandgap semiconductors, Silicon Carbide (SiC) and Diamond (3.2 and 5.5 eV bandgap at 300 K respectively) are of great interest due to their properties and applications (high-power, high-frequency, high-temperature and harsh environments). Indeed, SiC with its 4H polytype and Diamond are characterized by their high breakdown field, energy threshold of defect formation and thermal conductivity. Previous works were carried out between Aix-Marseille University and the CEA for these semiconductor detectors within the framework of their joint laboratory LIMMEX. Neutron fluxes were measured in the Zero-Power Reactor MINERVE at CEA (France, Cadarache) with a total neutron flux of 9.4×10^8 n/(cm2·s).
At present, a new objective is being targeted: the optimization of SiC-based diodes from the detector to the acquisition chain to perform online measurements under high neutron fluxes around 5.5×10^14 n/(cm2·s) (En > 1 MeV) as expected in the Jules Horowitz Reactor under construction at the CEA Cadarache center. To reach this aim, a step by step approach with intermediate neutron flux measurements is applied. The paper will present the results obtained simultaneously with a SiC-based diode and a Diamond detector in an ex-core irradiation channel (Tangential Channel) of the JSI TRIGA Mark II research reactor during an irradiation campaign in November 2021. The studied SiC-based detector is a p+n diode with a 21 µm sensitive thickness (1 µm 10^19 1/cm^3-doped p+-type and 20 µm 2×10^14 1/cm^3-doped n-type layers), a surface of 5.97 mm2 and a Boron-10 Neutron Converter Layer (NCL) for thermal neutron measurements. The Diamond detector from CIVIDEC Instrumentation corresponds to a single crystal detector elaborated by Chemical Vapor Deposition (sCVD) with a 140 µm thickness, a surface of 16 mm2 and a Lithium-6 NCL.
The first part of the paper will deal with the experimental set-up: the detectors, the detector holder, the conditioning and acquisition chains. The second part will present the parametrical study of the response of the SiC sensor. Several parameters were tested such as the applied bias voltage, the reactor power (from 1 to 1000 W) and therefore the neutron fluxes (from 10^7 to 10^10 n/(cm2·s)) and fluence. The detector pulses will be analyzed in terms of shape, amplitude, rise and decay times, and full-width at half-maximum before signal and data processing. The repeatability of the measurements will then be shown. Moreover, the influence of the bias voltage will be given on the count rate and on the Pulse Height Spectra (PHS). Moreover, the response of the diode as a function of the reactor power will be presented and discussed. The last part will be dedicated to the comparison between the responses of the SiC and Diamond detectors.

13.09.2022 09:30 Research reactors and radiation measurement

Research reactors and radiation measurements – 302

Radioactive environment characterisation using multi detector arrays and a 3-dimensional scanning Lidar

Matthew Ryan Tucker

Polymer Group, H.H. Wills Physics Lab. University of Bristol, Tyndall Ave., Bristol BS8 1TL, United Kingdom


Radiation mapping is a key part of both routine monitoring and decommissioning of nuclear environments, to ensure the safety of workers and that the reactor is operating as expected. Detected gamma radiation can have many sources: a relative homogenous distribution of radioactive material, shine paths where radiation from a source is partially attenuated, or hot particles, where small and highly active pieces of radioactive waste are unevenly distributed. Radiation sensors can be carried by workers, positioned around the facility or mounted on robots. When carried by a robot or a worker, the location of each measurement must be worked out, which can be done using Lidar sensors for Simultaneous Localisation and Mapping (SLAM). Using modern Lidar sensors, a 3-d dimensional model of the space can be made, and the radiation readings located within it.
In this paper the effectiveness of a radiation mapping system made up of a handheld Lidar scanning unit and 4 radiation detectors for radiation mapping and area characterisation is demonstrated. The detectors are evenly mounted on a belt, which is worn by a human operator, and is suitable for deployment in low dose environments. The body of the operator will attenuate the gamma rays by a percentage, which provides added directionality to the measurements. By using the experimentally derived response function of this 4 detector system, more information can be gained about the environment.
The added precision of this multi detector array can allow hotspots to be precisely located, but also allows tools to be used which can automatically differentiate between the 3 types of radiation source mentioned above. Taking this decision out of the hands of human operators improves the repeatability and consistency of these judgments.

The data threshold for making these judgments is explored, as providing real time information allows the operator to alter their mapping strategy and decision making accordingly.
This information can then be presented in the form of an automatically generated report on the environment, alongside a 3 dimensional point cloud with radiation readings shown in the form of coloured cubes. The advantage of this approach is that the interactive point cloud can be easily interpreted by workers and stakeholders familiar with the plant, instead of needing to refer to building schematics.

13.09.2022 09:50 Research reactors and radiation measurement

Research reactors and radiation measurements – 303

Ageing Management Program revision for the Pavia TRIGA MK II research reactor

Andrea Gandini1, Federico Alfinito2, Daniele Alloni1, Michelangelo Giordano1, Andrea Salvini1

1Laboratorio Energia Nucleare Applicata Universita degli Studi di Pavia, Via Aselli 41, 27100 – Pavia, Italy

2University of Pavia Laboratorio Energia Nucleare Applicata, Via Gaspare Aselli, 41, 27100 Pavia PV, Italy


The Laboratory of Applied Nuclear Energy (LENA) is an interdepartmental service centre of the University of Pavia in which, since 1965, a 250 kW TRIGA Mark II research reactor is in operation. Since 2014, the centre, implemented an ageing management program, that was submitted to the Italian regulatory body and finally approved in 2019 and the result was a combination of the already existing management system based on iso 9001 requirements (implemented in 2010) with ageing concepts as the ageing mechanism and safety classification of structures, system, and components. During the last few years new legal obligations needed a revision phase that started on two main cornerstones: continuation of safety culture promotion and the definition of a strategic plan for the practical control of ageing. This phase was done to manage all the practical aspects of ageing and to verify the compliance with IAEA and mandatory requirements. In this paper, a description of the revision process and both the activities carried out and those to be planned, will be provided.

13.09.2022 10:10 Research reactors and radiation measurement

Research reactors and radiation measurements – 304

The European Nuclear Experimental Educational Platform – ENEEP: Overview and Demonstration Activities

Vladimir Raduloviæ1, Anže Jazbec2, Luka Snoj1, Ján Hašèík3, Branislav Vrban3, Štefan Èerba3, Jakub Lüley3, Lubomir Sklenka4, Marcel Miglierini4, Ondrej Novak5, Helmuth Boeck Prof.Dr.6, Marcella Cagnazzo7, Mario Villa7, Szabolcs Czifrus8, Attila Tormási8

1Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Jožef Stefan Institute, Reactor Infrastructure Centre, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

3Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovièova 3, 812 19 Bratislava 1, Slovakia

4Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors, V. Holesovickach 2, 18000 Praha 8, Czech Republic

5Czech Technical University in Prague, Jugoslávských partyzánù 1518/3, 160 00 Prague 6 – Dejvice, Czech Republic

6TU Wien-Atominstitut, Stadionallee 2, A-1020 Wien, Austria

7Technical University Vienna, Atominstitut, Stadionallee 2, 1020 Vienna, Austria

8Institute of Nuclear Techniques Budapest University of Technology and Economics, Muegyetem rkp. 9, H-1111 Budapest, Hungary


The European Nuclear Experimental Educational Platform – ENEEP is currently being established by five European educational and research organizations in the framework of a Horizon 2020 project, initiated in 2019. The ENEEP partner institutions are the Jožef Stefan Institute (JSI, Slovenia), the Slovak University of Technology in Bratislava (STU, Slovak Republic), the Czech Technical University in Prague (CTU, Czech Republic), Technische Universität Wien (TU Wien, Austria) and the Budapest University of Technology and Economics (BME, Hungary). ENEEP is intended as an open educational platform, offering experimental hands-on education activities at the ENEEP partner facilities, which are in need in order to maintain high education standards in the nuclear field.
ENEEP is being developed on the basis on a comprehensive set of experiments performed at the ENEEP partner facilities, comprising the basics in Reactor Physics and Nuclear Engineering curricula, as well as more specific experiments focusing on particular aspects – investigated phenomena, types and working principles of detectors, etc. Novel education activities will be introduced and implemented in ENEEP, following scientific development in nuclear science and technology and nuclear instrumentation detectors stemming from research activities. ENEEP education activities will be offered in different formats (group and individual) and are targeted at university students at all educational levels and young professionals in the nuclear field.
This paper provides an overview of the ENEEP platform, focusing in particular on a series of demonstration courses, which was successfully carried out at the ENEEP partner facilities in early 2022, attended by university students from the EU and other eligible countries.

13.09.2022 10:50 Thermal hydraulics

Thermal-hydraulics – 401

SURET is a new form of subchannel thermohydraulic calculations

Áron Vécsi1, Gábor Házi2, Csaba Horváth3

1Centre for Energy Research, Konkoly Thege ut 29-33, Budapest-1121, Hungary

2Centre for Energy Research, Hungarian Academy of Sciences , Konkoly Thege M. út 29-33, H-1121, Hungary

3Centre for Energy Research Hungarian Academy of Sciences, P.O.Box 49, H-1525 Budapest, Hungary


SURET (SUbchannel REactor Thermohydraulics) subchannel analysis code has been developed by the Centre for Energy Research to simulate the behavior of mixing vane which was introduced in the new type of fuel assembly at Paks Nuclear Power Plant. The new fuel rods and its cladding are thinner than the ones used before and some of the spacer grids have been supplemented by mixing vanes to intensify the mixing in the assembly. SURET has been developed based on COBRA 3c. Their calculating modules are similar but slightly different energy equation is solved in SURET reducing significantly the computational time. SURET was originally developed for offline calculations. After performing some tests it became clear that it is suitable for online monitoring applications, too. For online calculations we applied more efficient algorithms for matrix inversion, optimizing the so called inner calculations. With these changes, we could significantly speed up the calculations (0.3 sec for the overall VVER-440 core) which was required for the application of SURET subchannel calculation in the online VERONA core-monitoring system. These modifications did not lead to any reduction of accuracy of the results computed. The new approach has already been integrated into VERONA and commissioned in the 2nd unit of Paks NPP. We have to install the new approach in the other units of the plant before the upcoming campaigns. We also designed a new graphical interface for SURET to support the user’s offline calculations, especially during the input preparation process and evaluation of results. The implementation of this new interface is an ongoing process.

13.09.2022 11:10 Thermal hydraulics

Thermal-hydraulics – 402

Thermal-Hydraulic Analysis of the 2nd Stage Hydroaccumulators Impact in LOCA Sequences with SBO

Elena Redondo-Valero1, César Queral2, Victor Hugo Sanchez-Espinoza3, Kevin Fernández-Cosials1

1Universidad Politécnica de Madrid, Alenza 4, 28003, Madrid, Spain

2Universidad Politécnica de Madrid (UPM), Nuclear Engineering Department, José Gutiérrez Abascal 2, 28006 Madrid, Spain

3Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany


VVER are one of the most common reactor types in the world. Moreover, a significant percentage of the Gen-III/Gen III+ reactors that are currently being built or have recently come into operation are VVER. Therefore, there is a growing interest in studying their behavior under both anticipated and accidental transients.

These advanced reactors have improved their safety systems to prevent core damage and ensure reactor integrity by implementing passive safety systems that do not require human actions or external power sources.

A joint effort between the Universidad Politécnica de Madrid (UPM) and the Karlsruhe Institute of Technology (KIT) has been made, within the ISASMORE project, in order to develop an integral plant model of a VVER-1000/V-320 reactor for TRACEp5 code.

In this work, the aim is to analyze the impact of implementing in the VVER-1000/V-320 model, the passive 2nd stage hydroaccumulators (HA-2) system, present in some VVER Gen III/III+ reactor designs. For this purpose, a sequence in which the VVER-1000 (without HA-2) design quickly reaches core damage is studied: a LOCA (Loss of Coolant Accident) along with an SBO (Station BlackOut).

13.09.2022 11:30 Thermal hydraulics

Thermal-hydraulics – 403

CFD Simulation of a VVER-1000/320 at Nominal Operating Conditions

Ossama Halim1, Andrea Pucciarelli2, Nicola Forgione3

1Universita di Pisa – Dipartimento di Ingegneria Civile e Industriale, Largo Lucio Lazzarino, 56122 Pisa, Italy

2Universita di Pisa, Dipartimento di Ingegneria Civile e Industriale, Largo Lucio Lazzarino 2, 56126 Pisa, Italy

3Universita degli Studi di Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione, Largo Lucio Lazzerino 1, 56100 Pisa, Italy


A full-scale STAR-CCM+ model for VVER-1000/320 application purposes was developed in order to predict the core outlet temperature distribution, the pressure losses experienced at different locations and to investigate the mixing coefficients between loops, for in-vessel flow. The research activity was carried out in the framework of CAMIVVER project: “Code And Methods Improvements for VVER comprehensive safety assessment project”. The primary aim of this work is to compare the measured and calculated core outlet temperature and mixing coefficient distributions at nominal operating conditions assessing the predicting capabilities of some selected turbulence models. The developed geometry consists of inlet nozzles, downcomer, lower plenum, core region, upper plenum, and outlet nozzles. The numerical simulations were performed using a computational grid of approximately 27.7 million polyhedral unstructured cells. The reference design of Kozloduy Unit 6 nuclear power plant was taken into account; with respect to the actual geometry of the vessel and its internals some simplifications were established in order to reduce the computational cost. Consequently, some regions were modelled as porous media, such as the core region, core basket, upper core plate, perforated barrel section and so forth. Also, additional pressure loss coefficients were imposed in the porous regions to reproduce the design pressure losses measured at the reference locations of Kozloduy-6 NPP. The CFD results predicted the presence of an azimuthal asymmetry of the loop flow centers relative to the cold leg axes, which is also observed in the experimental data. The azimuthal asymmetry shift is affected by the adopted turbulence model. Also, the distribution of the mixing coefficients at the fuel assemblies’ outlet slightly differs based on the adopted turbulence model. The average values of the core outlet temperature distribution in the calculation are in the same range of the measured plant data. Overall, the results show a good agreement with the corresponding average plant measured parameters and provide a better understanding of the involved phenomena. The promising results obtained in the frame of the present work represent a valuable benchmark showing the capabilities of the adopted numerical approach; the reliability of the adopted model will be thus further assessed against transient operating conditions in the frame of future applications to be included in the CAMIVVER project.

13.09.2022 11:50 Thermal hydraulics

Thermal-hydraulics – 404

Scaling down of PWR nuclear power plant secondary side for SIRIO experimental facility supported by system thermal-hydraulic codes

Samantha Larriba1, Rok Krpan2, Gonzalo Jimenez1, Elena Redondo1, César Queral3, Ivo Kljenak2

1Universidad Politécnica de Madrid, Alenza 4, 28003, Madrid, Spain

2Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia

3Universidad Politécnica de Madrid (UPM), Nuclear Engineering Department, José Gutiérrez Abascal 2, 28006 Madrid, Spain


In nuclear safety, experiments are being performed in experimental facilities to obtain useful information about the operation of real nuclear systems. However, experimental facilities are much smaller (usually at least by an order of magnitude) than the real systems the facilities are supposed to represent. For the experimental results to be applicable (in whatever way) to real systems, the operating conditions in real systems (or even in the entire nuclear power plants) have to be adequately scaled-down to the size of the experimental facilities. However, no universal principle has been accepted yet, so the scaling-down is usually performed using ad-hoc methods for specific cases.

Within the European project PIACE, a concept of Passive Isolation Condenser, in which the heat removal from the reactor core is slowed down by decreasing the steam condensation rate using injection of a non-condensable gas, has been proposed. In a Pressurized Water Reactor, the condenser should be connected to the reactor secondary side and limit the core cooling rate to reduce thermal stresses.

The suitability of the concept is verified in the SIRIO experimental facility located at the SIET company in Piacenza (Italy). First, scaled-down conditions to be applied in the experiment had to be determined. Second, simulations of the experiment had to be performed before the execution of the experiment itself. Although the results of the simulations are not expected to provide necessarily the same results (within uncertainty limits) as the results of the planned experiment, they might still be expected to support the adequacy of the prescribed scaled-down conditions.

To perform these simulations, JSI and UPM have developed models of the SIRIO experimental facility for the RELAP5 and TRACE thermal-hydraulic system codes, respectively. Several simulations were performed to test the capabilities of both codes to simulate the natural convection occurring in the closed loop. First, a steady state, where the steam flow is bypassed through a heat exchanger, with constant removed power, is established. The results of the steady state simulation are then used as initial conditions for the simulation of a transient with decreasing removed power.

The steady state and the transient results obtained by both codes are compared with the focus on the heat transfer in the heat exchanger pool and the heat losses in the pipes. Due to different code functions, the method of prescribing the temperature of the electrically-heated molten salts, which simulate the core power generation in the bayonet tube steam generator, (where the power to be removed is generated) is also discussed. From these comparisons that strengthen the confidence in the adequacy of the results, the differences between the codes are identified and common conclusions applicable for future natural convection facilities models are drawn. Last but not least, the adequacy of the scaling-down procedure is discussed.

13.09.2022 12:10 Thermal hydraulics

Thermal-hydraulics – 405

Multiphysics Analysis of an in-core Fission Product Removal System for the Molten Salt Fast Reactor

Federico Scioscioli1, Antonio Cammi2, Stefano Lorenzi2

1Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milano, Italy

2Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy


A foreseen feature of the Molten Salt Fast Reactor (MSFR) is the adoption of a bubbling system for the removal of gaseous and metallic fission products (FPs) consisting in the injection of Helium bubbles into the core and their extraction from the top of the fuel circuit. Bubbles are expected to remove FPs from the salt through various mechanisms, in particular, floating for metallic FPs and mass transfer for gaseous ones.

The present work is aimed at starting a comprehensive analysis on the He bubbling system, focusing on Gaseous Fission Products (GFPs) production, transport and removal. In particular, we investigate both its operational and its safety-related features, in order to get information useful for the design of such a system and to assess the convenience of its adoption. In order to perform the above analyses, we add the capability to simulate production, transport, and mass transfer of an arbitrary number of GFPs to a preexisting multiphysics solver, built with the OpenFOAM suite. Information on mass transfer is required in the form of a correlation for the Sherwood number and a value for the Henry coefficient. Previously, only Xe-135 had been considered for the MSFR analysis. While this isotope is certainly the most important poison in a thermal reactor, this is not at all the case in a fast environment like the MSFR.

The solver is then used to analyze the bubbling system and its impact on the safety of the reactor. As for the bubbling system characterization, the main figure of merit of the efficiency of GFP removal is the quantification of a characteristic removal time. In addition to that, the analysis of the bubbling system of the MSFR includes the evaluation of the poisoning effect, the activity and decay heat of the removed gas. The latter is an aspect crucial for the design of the off-gas unit since it require a dedicated cooling system, as shown by our results.

Among the safety-related studies, the developed multiphysics tool allow evaluating the void coefficient, determining upper bounds on the He flow-rate in order to avoid prompt supercriticality in case of loss of He injection. In addition, two different possible accidents are evaluated involving a complete loss of He injection, and complete loss of He removal. Results show the relevance of the thermal-hydraulics behavior in preventing prompt supercriticality in case of loss of He injection.

13.09.2022 14:00 Reactor physics

Reactor physics – 501

The Definition of Mini Labyrinth Benchmark for Radiation Shielding Calculations

Branislav Vrban1, Štefan Èerba1, Jakub Lüley1, Vendula Filova2, Vladimir Neèas1

1Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovièova 3, 812 19 Bratislava 1, Slovakia

2Slovak University of Technology in Bratislava Faculty of Electrical Engineering and Information Technology Institute of Nuclear and Physical Engineering, Ilkovièova 3, 81219 Bratislava, Slovakia


The Mini Labyrinth experiment is a neutron and gamma shielding experiment constructed at the Slovak University of Technology, Bratislava (STU). The STU Mini Labyrinth consists of NEUTRONSTOP shielding blocks, blocks of moderators, various neutron sources, and a graphite prism. This paper gives the precise definition of the Mini Labyrinth experiment which enables its modelling in the state-of-art transport codes and presents the newest experimental results of neutron and gamma field measurements. Various neutron and gamma detectors are used for the measurement including the Thermo Scientific RadEye portable survey meter, the SNM-11 BF3-filled corona detector, and the CR-39 track detectors, and the He3 tube detector. The first computational results are also presented, where the cross-section-induced uncertainties to the results are also assessed.

13.09.2022 14:20 Reactor physics

Reactor physics – 512

Uranium Mononitride fuel for the SMART reactor

Khurram Mehboob

King Saud University, Sustainable Energy Technologies Center, P.O.Box 800, Riyadh 11421, Saudi Arabia


The neutronics performance and safety characteristics of Uranium mononitride (UN) fuel in Small and Modular Reactor (SMR) System-Integrated Modular Advanced Reactor (SMART) were investigated to discern the potential non-proliferation, waste, and accident tolerance benefits that can be obtained from UN fuel. This work presents results from an initial evaluation of UN fuel in normal operating conditions of SMART reactor using OpenMC and compared their neutronics performance with UO2 in terms of fuel cycle length, reactivity coefficients, Fuel depletion (burnup), thermal flux, and fission product activity at the end of the fuel cycle length by keeping the UN fuel enrichment identical to the reference fuel (UO2). Results show that UN fuel can be operated beyond the designed length of the fuel cycle of the SMART reactor, which results in access to the positive reactivity at the end of the cycle about 4625.976 pcm, where the UO2 dropped to negative reactivity after three years. The total fission product radioactivity at the end of 3.5 years for UO2 and the UN has been founded as 1.003×1020 Bq and 1.023×1020 Bq, respectively.

13.09.2022 14:40 Reactor physics

Reactor physics – 503


Arturo Vivancos

Universitat Politecnica de Valencia, Camino de vera, s/n, 46022, Spain


Precise, effective, and optimised computer tools are required in the endeavour of simulating and predicting neutron phenomena taking place in a nuclear reactor core. Isotope evolution is a key aspect of reactor analysis and design. Historically, transmutation and disintegration (depletion or burnup) phenomena have been studied by simplifying the system complexity and reducing burnup chains.
This work’s main objective is to produce a computer program capable of performing precise burnup calculations considering a wide isotope species number and transmutation processes. In this regard, the work parts from a preliminary version, coupled to VALKIN-FVM-Sn deterministic transport code developed at UPV. This version is analysed in detail and improved, increasing its capabilities and reducing its computation times. Several ODE methods are compared, and the program is automated to perform fuel assemblies’ depletion calculations.
In this work, an initial fuel depletion program coupled with a transport code has been improved and depurated. Parting from satisfying results, the calculation process, execution time and performance have been enhanced. The response to modifications in the most influential computation and modelling parameters has been assessed. Different ODE solver methods were also compared. These efforts have resulted in a burnup code capable of predicting fuel assembly’s nuclide evolution with great precision and detail (high number of isotopes and transmutation processes), all of this making use of a reasonable computation time and computer resources.

13.09.2022 15:00 Reactor physics

Reactor physics – 504

Implementation and validation of the steady state SP3 approximation in the GRS FENNECS code

Silvia lo Muzio1, Armin Seubert2

1Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Forschungszentrum, Boltzmannstr. 14 , 85748 Garching, Germany

2Gesellschaft für Anlagen- und Reaktorsicherheit gGmbH, Boltzmannstr. 14, 85748 Garching bei München , Germany


From the growing interest in small modular reactors (SMRs) and micro modular reactors (MMRs), it arises the necessity to find a proper neutronic calculation tool for their safety assessment. Most of the deterministic neutronic solvers rely on the diffusion approximation that is derived assuming isotropic scattering, low probability of neutron absorption compared to scattering, as well as a weakly varying neutron flux in space. This last assumption may not hold for small cores, like the ones of SMRs and MMRs: here, due to their reduced size, high neutron flux gradients are present. An alternative to the use of the diffusion equation is the application of the third order Simplified Spherical Harmonics (SP3) approximation of the neutron transport equation, which is expected to perform better for SMRs and MMRs.
For this reason, the Finite ElemeNt NeutroniCS (FENNECS) code, currently under development at GRS, which already provides a diffusion solver, was expanded by a steady state SP3 solver. FENNECS offers a high geometrical flexibility, which is essential to model complex systems like SMRs and MMRs.
In this paper, starting from the transport equation, the steady state SP3 approximation of the neutron transport equation is derived. Then, in order to implement the SP3 equations in the FENNECS code, which is based on the Finite Element Method, these are casted into the Galerkin (weak) form. Finally, based on benchmark exercises, the correct functionality of the SP3 solver implemented in FENNECS is shown.

13.09.2022 15:20 Reactor physics

Reactor physics – 505

Preliminary Neutronics Analysis of APR1400 Core Loaded with U3Si2-FeCrAl Accident-Tolerant Fuel

Khawla Ali Alhammadi1, Fawzeya Bin Tamim1, Amna Khalili1, Donny Hartanto2, Iyad Al Qasir1

1University of Sharjah, P.O. Box, 27272 Sharjah, United Arab Emirates

2Oak Ridge National Laboratory, P.O.Box 2008, Oak Ridge, Tennessee 37831-6162, USA-Tennessee


This study presents the preliminary investigation of an accident-tolerant fuel in the Barakah nuclear power plant (BNPP). BNPP is located in the United Arab Emirates, and it consists of four Gen-III advanced pressurized water reactor (APR1400) that use uranium dioxide fuel (UO2) contained in a zircaloy (Zr) tube. After the Fukushima Daiichi accident, the feasibility of accident-tolerant fuels (ATF) has been widely investigated to improve fuel performance and increase reactor safety. In this study, a proven nuclear fuel such as the high-density uranium silicide (U3Si2) and improved cladding such as FeCrAl alloy are implemented into BNPP. Several important neutronics parameters at the equilibrium core are evaluated and compared with the current UO2/Zr fuel system. The equilibrium core is achieved by simulating multiple cycles, including fuel shuffling and refueling. For this purpose, Monte Carlo Serpent 2 code, in conjunction with the latest nuclear data library, ENDF/B-VIII.0, is used. The parameters of interest include the minimum enrichment required for the ATF to maintain criticality for the same cycle length, the temperature reactivity feedback coefficients, control rod worth, and power profiles in the core. Finally, the discharged fuel’s activity and decay heat are also analyzed.

13.09.2022 15:40 Poster session 1

Education and training – 204

Nuclear Competence Building via Education and Training Initiatives: activities of the SCK CEN Academy

Michele Coeck, Clarijs Tom, Niels Belmans

CEN/Serma – Lepp, BAT.470, 91191 GIF-SUR-YVETTE, France


Preserving and extending nuclear knowledge on fundamental and peaceful applications of ionizing radiation to serve society, is one of the key elements in SCK CEN’s research policy. Thanks to its thorough experience in the field of nuclear science and technology, its innovative research and the availability of large and unique nuclear installations, SCK CEN is an important partner for education and training in Belgium as well as at international level.
In order to maintain and extend a competent workforce in nuclear industry, healthcare, research, and governmental organizations, and to transfer nuclear knowledge to the next generations, the mission of the SCK CEN Academy comprises (i) guidance for young researchers, (ii) organisation of courses, (iii) policy support with regard to education and training matters and (iv) caring for critical-intellectual capacities.
It coordinates the knowledge and competence building actions on various nuclear topics SCK CEN is performing research on, such as nuclear materials science, reactor engineering, radiation protection, nuclear safety, emergency management, decommissioning, waste and disposal, etc.
At academic and research level, scientists at SCK CEN collaborate via its Academy with universities to hosts several PhD students enhancing new findings in support of nuclear science and applications. Our experts are also available to mentor Bachelor and Master students and supervise their thesis or internship. Moreover, the SCK CEN Academy supports high school teachers with educational materials and likes to immerse pupils and the general public into the fascinating world of nuclear science and technology.
We also provide academic education on various topics related to nuclear applications: nuclear engineering, reactor physics, radiation protection, nuclear safety, materials sciences, nuclear fusion, radioactive waste management, dismantling & decommissioning and nuclear technology assessment. In this framework, guest lectures, practical demonstrations and technical visits related to various nuclear topics are embedded in academic programmes for future experts in nuclear engineering, nuclear safety and radiation protection. For professionals working in the nuclear sector, the SCK CEN Academy organizes several training courses, in many cases customized and in collaboration with external international experts.
Thanks to networking and participation in international programmes, the SCK CEN Academy can also contribute to a better harmonization of education, training practice and skills recognition on an international level. In this way we are partner in various European funded projects on nuclear education and training, and are affiliated in the most prominent nuclear networks and associations.
Understanding the benefits and risks of radioactivity including nuclear applications requires scientific and technical insight and training, but also an insight in the context and a sense for the societal and philosophical aspects of the situation. The SCK CEN Academy is committed to encourage a critical mind and objectivity among students, trainees and PhD researchers in the nuclear domain.
This presentation will highlight the initiatives of the SCK CEN Academy and will show how cooperation with several stakeholders like universities and industry, also at international level, contributes to a more efficient transfer of knowledge, skills and competences in nuclear sciences and technology.

13.09.2022 15:40 Poster session 1

Education and training – 206

Nuclear Education – what influence does online teaching have – a cause study in Austria

Milena Zehetner1, Eileen Langegger2, Helmuth Boeck Prof.Dr.3

1Osterreichische Kerntechnische Gesellschaft (Austrian Nuclear Society) Atominstitut, Stadionallee 2, A-1020 Vienna, Austria

2Austrian Nuclear Society, Rudolf Zöllner Strasse 31, 2500 Baden, Austria

3TU Wien-Atominstitut, Stadionallee 2, A-1020 Wien, Austria


The strong increase in the student numbers in nuclear subjects at Austrian Universities during the Pandemia triggered the following research questions.

The last two years presented universities with big challenges. Conventional teaching methods had to be adapted and new approaches developed. In respect to Austria’s nuclear education, the adjustment to the pandemic situation was mastered with the experience of the Young Generation Network (YGN).

Courses were, if feasible, shifted towards online formats. The new approach was well accepted by students. Nuclear lectures and courses experienced a high number of participants during the online teaching format and also the assigned project- and bachelor-theses increased compared to the time before covid.

For evaluating this strong rise, a survey was handed out to students of which the results are presented in this paper. It also explains the teachers’ as well as the students’ point of view and challenges both were facing during that time and how this information can be used to maintain this positive development in Austria’s nuclear education.

It also looks at some general questions on the opinion of the students, and how the lectures have influenced their opinion or deepened their knowledge.

13.09.2022 15:40 Poster session 1

Education and training – 207

Training and Tutoring for the Nuclear Safety Experts of Non-EU Countries

Tamás Pázmándi1, Giovanni Bruna2, Csilla Pesznyák3, Gerard Cognet4, Alessandro Petruzzi5, Gabriel Pavel6, Márton Benke7, Branislav Hatala8, Elektra Tsigaridas9, Dorottya Jakab1

1Centre for Energy Research, Konkoly Thege ut 29-33, Budapest-1121, Hungary

2NucAdvisor, 168/172, boulevard de Verdun, Energy Park – Building 4, 92408 COURBEVOIE CEDEX, France

3Budapest University of Technology and Economics, Mûegyetem rkp 3-9, Budapest 1111, Hungary

4CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France

5University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Livornese 1291, 56122 Pisa, Italy

6European Nuclear Education Network, Rue d’Egmont 11, 1000 Brussels, Belgium

7University of Miskolc Department of Analytical Chemistry, Izpolni naslov!, 3515 MISKOLC – EGYETEMVÁROS, Hungary

8VÚJE Trnava Engineering, Design and Research Organization, Ltd., Okružna 5, 91864 Trnava, Slovakia

9European Commission, Rue Montoyer 75, B-1049 Brussels, Belgium


Safe utilization of nuclear energy requires competent, independent, and adequately financed National Nuclear Regulatory Authorities (NRAs) and Technical Support Organizations (TSOs). Because of the high demands on technical competence, the continuous availability of new information (development of new reactor types, new safety mechanisms or new assessment methodologies), the recruitment of new staff, there is always a need for general, in-depth and specific training for the experts of NRAs and TSOs to build and maintain their necessary knowledge and skills. The European Union (EU) supports the achievement of the above in countries outside the EU through the European Instrument for International Nuclear Safety Cooperation (INSC) and has initiated several actions to provide training for countries in need of technical assistance.
Training & Tutoring initiative to support competence building worldwide is part of the INSC’s efforts towards making the EU a global reference in matters of nuclear safety and radiation protection, emergency preparedness and regulatory framework.
Phase 5 of the European Commission’s INSC project has been launched in January 2022. It is implemented by a Consortium led by EK (Hungary), having members of NucAdvisor (France), N.IN.E. S.r.l. (Italy), VUJE, a. s. (Slovakia), Uni-Energy Ltd. (Hungary) and ENEN (Belgium). Throughout the nearly three years of the project, several courses – both in the form of trainings and several weeks tutoring – and assistance will be provided for the experts of non-European countries’ NRA(s) and TSO(s) to strengthen their capabilities with regard to their tasks and responsibilities related to radiation protection and nuclear safety. Developing such expertise is more than a matter of education, as it involves not only the transfer of technical knowledge, experiences, and best practices, but also helps promoting the European nuclear safety culture.
The programme of the current project will be presented.

13.09.2022 15:40 Poster session 1

Education and training – 208

Nuclear Technology Courses in Nuclear Training Centre Ljubljana

Tomaž Skobe

Jožef Stefan Institute, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The paper presents experiences from performing nuclear technology courses at Nuclear Training Centre Ljubljana. There are two types of important courses, conducted for NPP Krško staff and other organizations, dealing with nuclear technology. The first course is called NPP Technology (the acronym in Slovenian language is TJE) and is intended for future control room operators. This course is the first, theoretical part of the initial training of licensed operators (later stages – NPP systems and simulator training – take place at the location of the NPP). Approximately 5 months are devoted to different topics, such as nuclear and reactor physics, thermal-hydraulics and heat transfer, radiation protection, electrical engineering, materials, and nuclear safety.
The second course, Basics of Nuclear Technology (in Slovenian OTJE) is suitable for other NPP technical personnel, technical support organizations, regulatory body, etc. In 2022 the 43rd edition of the course was conducted. This course consists of two parts: theory (4,5 weeks) and NPP Systems (3,5 weeks).
The paper will present the course organization, materials preparation, preparation and supervision of lectures, and feedback from participants.

13.09.2022 15:40 Poster session 1

Research reactors and radiation measurements – 305

Preparation of a New Research Nuclear Reactor in Slovenia

Jan Malec1, Vladimir Raduloviæ2, Mitja Uršiè3, Iztok Tiselj3, Borut Smodiš4, Klemen Ambrožiè2, Anze Pungercic2, Christophe Destouches5, Robert Jacqmin6, Gilles Bignan7, Xavier Wohleber8, Luka Snoj2

1Jožef Stefan Institute, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

3Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia

4Institut “Jožef Stefan”, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

5Commissariat a l’Energie Atomique – Centre d’Etudes de Cadarache DRN/DER, Bat 238, 13108 St Paul Lez Durance Cédex, France

6Commissariat a l’Energie Atomique et aux Energies Alternatives, Batiment Le Ponant D 25 Rue Leblanc , 75015 Paris, France

7CEA France, CEN Saclay ORE/SRO, France

8CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France


In light of climate change and the global initiative to transition to carbon neutrality, the European Union has recently recognized nuclear energy as sustainable in the Taxonomy Directive. Several member states and the United Kingdom announced plans to expand their nuclear power fleets. In support to this nuclear program, the European fleet of research reactors is aging. Therefore, there is a clear need for new research reactors that would meet the demand for nuclear research and technologies in the coming decades through 2100.

The Jožef Stefan Institute, in collaboration with the French Alternative Energies and Atomic Energy Commission CEA, has initiated activities to prepare a comprehensive feasibility study report. We have compiled a list of potential stakeholders, selected reactor technology, and developed a preliminary timeline for moving toward a new research reactor. It is likely that –as many European Research Infrastructures, this new research reactor will be managed as an international consortium project with many stakeholders. Therefore, the technology must be selected to cover a wide range of scientific and technical user’s needs for the benefit of European Union Member States.

This new research reactor facility will consists in two reactor types: first one will be a pool-type reactor, cooled and moderated with light water and surrounded with a heavy water reflector and neutron beams for giving the scientific community high flux of neutron. In this way, we will be able to conduct research in support of the European fleet of existing and future nuclear power plants, including small modular reactors based on pressurized water reactor technology. Secondly, in addition, a water-cooled pool reactor at zero power has proven to be suitable for research and education because it provides easy access to the reactor core, is simple to operate, and is very flexible. To meet the need for high flux applications as well as the need for education and training and the performance of benchmark reactor physics experiments, the idea is to build a nuclear facility with two reactor cores. The first would be a multipurpose research reactor with a thermal power of a few megawatts. Such a facility would be used for neutron activation analyzes, radiation hardness studies, instrument testing and calibration, neutron radiography, neutron transmutation doping, radioisotope production, testing of new additive materials, and beam experiments with cold and fast neutrons. The second core would be a flexible and versatile zero-power reactor with a maximum power of several kilowatts and would allow benchmark experiments to be performed with different fuel types and neutron spectra.
The paper will present in detail the concept design and the first of list of potential stakeholders.

13.09.2022 15:40 Poster session 1

Research reactors and radiation measurements – 306

Dynamic Mode Decomposition Analysis of a Cooling Channel of the TRIGA Mark II Reactor

Carolina Introini1, Vittoria Brega1, Antonio Cammi2

1Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

2Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy


As nuclear reactors present unique features compared to conventional systems, identifying the best performing advanced modelling strategies remains an ongoing challenge from the point of view of accuracy and ef?ciency, especially for the safety aspect. In this context, Model Order Reduction (MOR) techniques offer a promising solution to the trade-off between solution accuracy and computational times, especially for multi-query scenarios. Traditional MOR techniques have been successfully applied in the nuclear engineering community to study the long-term behaviour of the system, however model-based MOR presents the computational bottleneck of needing evaluations of the full-order system in order to provide the data to build the reduced model.

Dynamic Mode Decomposition (DMD) is an equation-free MOR technique able to represent even complex models with explicit temporal dynamics based only on the observed data, without requiring any knowledge of the underlying governing equations. DMD allows the extraction of the time-varying characteristics of the system and of the governing dynamic structures from the available snapshots and, compared to other MOR methodologies, allows the evaluation of a low-dimensional surrogate of the dynamic matrix A, on which dynamic and stability analysis can be performed, and to predict the future system behaviour even without observations. As its application for nuclear-related applications is still not widespread, this work carries out an optimisation of the DMD algorithm for the reconstruction and prediction of reactor transients. The selected benchmark test case is a cooling channel of the TRIGA Mark II Reactor, with the aim of optimising the algorithm for its future application on the whole reactor system. The results show how, providing enough data are available at the beginning of the transient, DMD is able to correctly predict the pseudo steady-state behaviour of the system even in absence of data.

13.09.2022 15:40 Poster session 1

Research reactors and radiation measurements – 307

Multi-Technique Experimental Benchmark at the JSI TRIGA Reactor for the Modelling of Nuclear Instrumentation

Vladimir Raduloviæ1, Loic Barbot2, Damien Fourmentel3, Elsa Dupin4, Adrien Gruel5, Vincent Chaussonet4, Domergue Christophe5, Herve Philibert4, Clement Fausser4, Alexandre Subercaze4, Anze Pungercic1, Klemen Ambrožiè1, Ingrid Švajger1

1Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2CEA, DES, IRESNE, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache , F-13108 Saint-Paul-Lez-Durance, France

3CEA Cadarache, IPSN/DRS/SEMAR BAT. 700, 13108 St. Paul Lez Durance, France

4CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France

5CEA-Cadarache DEN/DER/SPEX/LDCI, Bâtiment 238 – Piece 10, F13108 Saint-Paul-lez-Durance, France


In the last decade, the utilization of Jožef Stefan Institute (JSI) TRIGA reactor as an experimental facility for research, development and experimental testing of nuclear instrumentation detectors has steadily been increasing. These activities are made possible by comprehensive past efforts on the characterization of the experimental locations available in the reactor, mostly performed in the framework of collaboration projects between the JSI and the French Atomic and Alternative Energies Commission (CEA), each covering a specific experimental technique.
In 2020 a new JSI-CEA collaboration project was launched with the aim of performing a multi-technique experimental benchmark at the JSI TRIGA reactor, to further support research and development in the field nuclear instrumentation. The experimental techniques employed involved measurements with miniature fission and ionization chambers (FC, IC), self-powered neutron and gamma detectors (SPND/SPGD), thermoluminiscent (TLD) detectors, calorimeters and neutron activation dosimetry. The experimental measurements were focused on obtaining information on spatial distributions of the neutron and gamma flux within the reactor core (FC, IC, TLD, neutron dosimetry) and information to support the determination of the neutron spectra (neutron dosimetry) by unfolding techniques. The experimental measurements were successfully carried out in two experimental campaigns at the JSI TRIGA reactor in March and May 2022.
In parallel to the experiments, modelling activities using Monte Carlo particle transport codes have been ongoing both at JSI and CEA. A detailed computational model of the JSI TRIGA reactor was created in TRIPOLI4, and verified against past benchmark experiments. The experimental measurements will be reproduced computationally both by JSI (MCNP, SERPENT) and CEA (TRIPOLI4) and compared, enabling further improvements in the modelling of the reactor and nuclear instrumentation responses.
This paper presents the preparation of the experimental measurements and reports some first results obtained from experiments and modelling.

13.09.2022 15:40 Poster session 1

Research reactors and radiation measurements – 308

Analysis of water activation loop at the JSI TRIGA research reactor

Domen Kotnik1, Anil Kumar Basavaraj2, Igor Lengar1

1Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Water as a primary coolant is used in the majority of fission reactors today and will also play an important role in the performance of fusion reactors. However, it is still poorly understood as a radiation source. After it is irradiated and activated, the cooling water flows through the cooling circuit, usually outside the primary biological shield surrounding the reactor vessel, dispersing radioactivity throughout the plant. The threshold energy for the main water activation reaction, i.e., 16O(n,p)16N, is about 10 MeV. Thus, neutrons in fusion reactors result in water activity that is 5 orders of magnitude higher than in fission reactors of similar power. Many computational analyses of the water activation process have been performed for ITER and DEMO. However, the results are subject to uncertainties and therefore of poor quality due to lack of experimental nuclear data, inaccurate computational methods/codes, and experimental facilities to validate the methodology experimentally.

With this in mind, a closed water activation loop is being constructed at the Jožef Stefan Institute (JSI) research reactor TRIGA Mark II that will serve as a well-defined and stable 6 MeV – 7 MeV gamma-ray source. The main focus of the work is to analyse different designs of the main irradiation part of the water activation loop, which is located inside the radial piercing port right next to the reactor core. The main design criteria are the effective water volume, pressure drop, flow velocity profile, and reaction rate map. Since the moving activated water is a time and spatially dependent radiation source, transport calculations must be coupled with CFD calculations. Similar conditions exist inside a water-cooled fission/fusion reactor.

The analyses performed will provide important details for the final design of the entire irradiation facility, since the design/shape of the irradiation part directly affects the overall activity that can be achieved with such an irradiation facility. The main objective is to perform water-activation based benchmark experiments.

13.09.2022 15:40 Poster session 1

Research reactors and radiation measurements – 309

Development of a Radioactive Particle Tracking Method for a Moving Radioactive Source with One Scintillator Detector

Roque Antonio Santos Torres, Verónica Bedón

Escuela Politecnica Nacional, , Ecuador


This research is the first step in a pioneer project in Ecuador for developing a low-cost Radioactive Particle Technique (RPT). The objective of this work was to test the ability of one sodium iodide (NaI) scintillation detector to obtain information about a moving radioactive particle for the posterior trajectory reconstruction. Other experiments use the same technique but use several detectors that were calibrated with static particles, one detector at a time. This research tests the concept of calibrating a detector with a moving particle. For this purpose, a movement system was built to move a 137Cs source of 10 µCi activity in a straight trajectory. The movement system allows for imparting two velocities to the source. These velocities were measured with two different means: a built-in encoder that measures the revolutions of the moving motor, and external laser sensors that determine the moment at which the source leaves its initial position and reaches its final point. Meanwhile, the scintillator detector was placed in different locations. The height from the floor to the detector, the detector’s location relative to the total length of the trajectory, and the distance from the source to the detector were changed to place the detector in 27 different positions. In this manner, the response of several detectors looking at the moving particle was simulated. Experimental data were obtained using a data acquisition system (DAQ NI 9219) directly connected to the NaI scintillator detector. The registered information was stored in form of voltage impulses as a function of time. Two computer codes were used to treat this raw data. The first one translates the matrix impulses versus time to a new matrix in the form of impulse counts versus position. This translation was done based on the highest voltage value that was considered to originate from the interaction with the radioactive source. The second code treated the resultant matrix with a Kernel methodology before applying a numerical derivation to reconstruct the source velocity and accelerations as it moves, thus using the scintillator detector as a velocity sensor. Measured velocity was compared to the nominal velocity at which the source was moved. Results of this research showed that for certain locations of the detector, there was a way to establish a direct relation of the impulse count with the location of the source. This relation can be used to reconstruct the position of the source when it is measured with a detector in the same location. Results also show that for different locations, there is a difference in the way this impulse count versus positions is registered. More research is needed to establish if this can be used to provide location sensitivity to the detector, but the results are promising. On the other hand, for other locations, the sensibility of the detector is not enough to differentiate the radiation coming from the source from that originated in the environment. In those locations where the detector has enough sensibility to register the radiation coming from the source, the difference between the source nominal velocity and calculated velocity was below 5 %.

13.09.2022 15:40 Poster session 1

Research reactors and radiation measurements – 310

Jožef Stefan Institute TRIGA Research Reactor Activities in the Period from September 2021 – August 2022

Anže Jazbec1, Sebastjan Rupnik1, Vladimir Raduloviæ2, Borut Smodiš3, Luka Snoj2

1Jožef Stefan Institute, Reactor Infrastructure Centre, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

3Institut “Jožef Stefan”, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The Jožef Stefan Institute (JSI) has been operating a 250 kW TRIGA research reactor since 1966. Safety performance indicators (SPI) have been monitored for over ten years. Examples of the monitored parameters are; the operating time, the number of irradiated samples, doses received by operating staff, and the activity of radioactive gases released into the environment. In the paper, SPIs for the year 2021 will be presented and analyzed. Such an analysis is an important tool to improve the future safe operation of the research reactor.
Furthermore, new research work carried out during the past 12 months will be presented. Several research campaigns resulted from a collaboration between CEA and JSI. For the first time, calorimetric measurements were done to evaluate gamma heating of different materials during reactor operation. In January, an extensive pulse campaign was carried out to evaluate the response of micro fission chambers. In less than two weeks, 150 pulses were done. Several campaigns were made to characterize our core and validate computer codes using micro fission chambers, thermoluminescence dosemeter and dedicated foils that activate during reactor operation. The work in the characterization of self-powered neutron detectors and irradiation of FT-TIMS capsules continued from previous years.
In the field of education, there were plenty of activities performed in the last year. Some exercises were still carried out remotely, depending on the Covid situation. We hosted students from various universities (University of Ljubljana, Uppsala University, Aix Marseille University and Politecnico di Milano). We hosted a demonstration course of ENEEP (Europen Nuclear Experimental Education Platform). The platform is aimed at students and young professionals in the nuclear field and provides them access to nuclear facilities for educational and research purposes. For the first time, an experimental reactor physics course was organized for the participants of the SARENA project. After just a one-year break, we hosted trainees from our Nuclear power plant Krško who attended Nuclear technology course. They performed seven practical exercises at the TRIGA reactor.
In summer 2022 we plan to perform a detailed inspection of the liners of reactor pool and spent fuel pool. Special attention will be given to the radial piercing beam port which we plan to use for the future experimental device – the water activation loop. In August, we plan to replace all the components of the secondary cooling loop that are located inside the reactor building.

13.09.2022 15:40 Poster session 1

Research reactors and radiation measurements – 311

Analysis of the void coefficient in Pavia TRIGA Mark-II reactor: Monte Carlo numerical evaluation and comparison with experimental data

Riccardo Boccelli, Antonio Cammi, Carolina Introini, Stefano Lorenzi

Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy


The Pavia Triga Mark-II is a research reactor designed by General Atomic aimed at being used for training and research purposes. The peculiar used of uranium zirconium hydride (UZRH) as fuel provides the reactor with a large prompt negative reactivity coefficient. This choice, along with the low nominal power (i.e., 250 kW) and the pool type configuration ensures an excellent level of passive safety. In this light, the evaluation of void formation on the reactivity is important, despite the formation of bubbles due to boiling is not foreseen in normal operation. On the other hand, subcooled boiling or formation due to leakages from fuel element and/or samples could be considered possible source of void.
The evaluation of reactivity void coefficients is not straightforward since depends on the balance between the opposite contribution of capture and scattering after a perturbation in the multiplication factor. For this reason, the void coefficient is also position dependent and extremely non-linear, depending on the real quantity of the void formation.
This work aims at providing an extensive analysis of the different mechanisms involved in the evaluation of void effect in the Triga Mark-II reactor installed at the Applied Nuclear Energy Laboratory (LENA) of University of Pavia. As reference, we take the experimental procedures employed for the evaluation of void coefficient to be reproduced and analyzed through the Monte Carlo code SERPENT. A model of the Pavia Triga Mark II reactor, previously developed with the SERPENT code, is employed in the analysis. It adopts fresh low-enriched fuel of type 101 and 103, arranged as in the latest reactor configuration. The model has been already validated against control rod calibration curves and neutron flux experimental data. The experiment analyzed consist in placing aluminum or polyethylene samples filled by air or water in the central channel of the reactor which is usually not filled with a fuel element and used for irradiation. In addition to the comparison of the experimental results, the analysis allows both identifying the different components of the void coefficient, perturbing the single cross section (total, elastic, capture, …) and evaluating the sensitivity coefficient to the multiplication factor. The results show that the void coefficient is dependent on the parameters that may affect the moderation ratio as the choice of the casing material, the amount of water/air inserted (i.e., the void fraction), the radial and axial position inside the core.

13.09.2022 15:40 Poster session 1

Research reactors and radiation measurements – 312

Testing of Silicon Carbide Neutron Detector for Detection of Fast Neutrons

Ylenia Žiber

University of Ljubljana Faculty of Mathematics and Physics , Jadranska 19, 1000 Ljubljana, Slovenia


As the supply of 3He diminishes, a need for neutron detectors based on other technologies than 3He has arisen. In recent years, semiconductor detectors have become popular, especially silicon carbide (SiC) detectors. Such detectors have been developed in the E-SiCure project and further optimized in E-SiCure2 project. Optimizations have been focused on scaling the detection efficiency to thermal neutrons and expanding the detection capabilities to other radiation types, in particular fast neutrons.
A computational study was carried out in search of new neutron converter materials for the detection of fast neutrons. Among the identified candidates, the converter material of choice was KCl, with two isotopes (39K and 35Cl) having a sufficiently high (n,p) reaction rate. Testing of the fast neutron converter material with SiC detectors was performed at the Jožef Stefan Institute TRIGA Mark II research reactor. In the experiments, a thermal neutron absorber (10B4C on Cu substrate) was mounted in front of the SiC detector to reduce the thermal neutron component as much as possible. From the measurements performed, a clear response to fast neutrons was observed even without the presence of converter material, attributable to recoil carbon and silicon nuclei.
This paper presents the preparation of fast neutron converters and the experimental testing of SiC detectors for fast neutron detection.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 406

Experimental study on bubble size distributions in horizontal narrow-gap annular heat exchanger

Boštjan Zajec, Leon Cizelj, Boštjan Končar

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Boiling is an effective heat transfer mechanism, commonly present in nuclear power plants and in other thermal engineering applications. Despite long history of boiling flow research, some underlying phenomena are still not fully understood. Bubbles change in size and shape as they move through the liquid, due to evaporation on the heated wall, condensation in the subcooled liquid, and interactions with other bubbles. This paper focuses on experimentally determining the bubble size distribution to capture the combined effect of these mechanisms. Boiling flow of refrigerant R245fa is studied in a temperature-controlled narrow-gap annular heat exchanger. Two different operational regimes are analyzed and visualized with a high-speed camera. Image processing with manual and neural-network bubble recognition is used to characterize bubbles and determine the bubble size distribution. Experimental setup, methods of experimental analysis and results are presented and discussed.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 407

Simulation of flow in PWR reactor pressure vessel downcomer

Aljaž Kekec1, Jure Marn2, Ivo Kljenak3

1University of Maribor Faculty of Mechanical Engineering, Smetanova ulica 17, 2000 Maribor, Slovenia

2Faculty of Mechanical Engineering, Aškerèeva 6, 1000 Ljubljana, Slovenia

3Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


In a Pressurized Water Reactor (PWR), coolant enters into the downcomer of the Reactor Pressure Vessel (RPV) through two, three or four cold legs, with inlets evenly spaced over the perimeter. The mixing of the flow in the downcomer determines the flow conditions (velocity and temperature fields) in the RPV lower plenum and further on in the reactor core. Unfortunately, as the installation of measuring devices in the downcomer would be impractical and costly, and would unnecessarily disturb the flow, the actual pattern of the flow in the downcomer is unknown. The knowledge of the flow pattern would offer additional insights into the phenomena in the RPV.

With the advent of Computational Fluid Dynamics (CFD), the flow in the downcomer may be simulated on the local instantaneous scale, providing a detailed picture of the flow. Although simulations probably do not replicate exactly the flow, the results may still be considered as a reasonable approximation of the actual flow.

The flow in the downcomer of a two-loop PWR RPV was simulated on the local instantaneous scale, using the CFD code CFX, both at normal operating and at break flow conditions. The flow was assumed to be isothermal, so the issue of pressurized thermal shock was not considered. The simulations provide insights into the velocity field in the downcomer. Furthermore, a simulation with obliquely (instead of perpendicularly) mounted cold legs was performed to evaluate whether such a modified design would be more favorable for the flow mixing.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 408

Simulation of liquid waves with flow reversal in stratified counter-current flow with a hybrid multi-fluid model

Matej Tekavèiè1, Richard Meller2, Benjamin Krull2, Fabian Schlegel2

1Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Institute of Fluid Dynamics, Helmholtz-Zentrum Dresden – Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden, Germany


Processes involving gas and liquid flows are important for reliable, efficient and safe operation of many industrial applications, such as electricity generation in nuclear power plants. Many different two-phase flow patterns can appear in these systems, with a wide range of scales considering both interfacial and turbulent structures. Stratified flow, i.e. phases being separated with a smooth or wavy interface, is one of the most important regimes for safety analyses.

The present paper presents simulations of an isothermal stratified counter-current flow of air and water in a rectangular channel of the WENKA experiment (Stäbler, T.D., 2007, PhD Thesis, Univ. Stuttgart). The partial flow reversal regime with liquid waves is considered. The wavy air-water surface is resolved with a hybrid multi-fluid model, featuring consistent momentum interpolation numerical scheme, partial elimination algorithm to handle strong drag coupling between phases, and interface sharpening method. The Unsteady Reynolds Averaged Navier-Stokes (URANS) approach with the k-? SST (Shear Stress Transport) model and interface turbulence damping is used to model the turbulent stratified flow with wavy surface. Simulations are performed with the open source C++ library OpenFOAM. Results are validated with experimental data for the height of liquid surface, profiles of velocity and turbulent kinetic energy, and the amount of reversed liquid flow.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 409

Thermal-hydraulic analysis of TEPLATOR moderator cooling system

Tomáš Koøínek1, Martin Lovecký2, Ondøej Burian2, Radek Škoda2

1Czech Technical University in Prague, Czech Institute of Informatics, Robotics and Cybernatics, Jugoslávských partyzánù 1580/3, 160 00 Prague, Czech Republic

2University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic


The heavy water reactor concept TEPLATOR contains separate independent systems for the primary coolant and the moderator. The present study analyses the low-pressure moderator cooling system of TEPLATOR during full-power operation. The moderator is heated from neutron thermalization, gamma rays absorption, fission product decay and decay of activation products. Additionally, heat transfer from the coolant channels has to be taken in the analyses of the moderator cooling system. Preliminary thermal-hydraulic analyses of the cooling system are supplemented by CFD simulations of heat and fluid flow in the moderator’s vessel. Results from CFD simulations are further assessed to evaluate and optimize the moderator cooling system. with particular attention to inlets’ and outlets’ locations.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 410

Passive Isolation Condenser Modeling With Apros Computer Code

Luka Štrubelj, Klemen Debelak

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia


A fully passive system for decay heat removal, based on the concept of isolation condenser is the subject of project PIACE. The feasibility study of passive isolation condenser application to several types of nuclear power reactors, such as: pressurizer power reactor, boiling water reactor, CANDU, lead cooled fast reactor and accelerator driven system MYRRHA was performed. This paper focuses on application of isolation condenser to pressurized water reactor. The reference power plant was defined. The station black out accident was identified as accident where isolation condenser can be applied if other decay removal systems fails. Numerical simulations of the primary system and isolation condenser were performed with computer code APROS. The results show that such passive isolation condenser is capable of removing decay heat.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 411

Parametric study of Population Balance Model on the DEBORA benchmark experiments

Aljoša Gajšek, Matej Tekavèiè, Boštjan Konèar

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Subcooled boiling is an important heat transfer mechanism that occurs in many industrial processes where high heat fluxes are involved. In the past, the phenomena of heat transfer by boiling up to critical heat flux, where heat transfer is greatly reduced and resulted in dramatic temperature rise, has been modelled with empirical correlations limited to specific coolant, geometry, and range of operating conditions. Due to continuous development of three-dimensional multi-phase computational fluid dynamics modelling capabilities, the simulation of complex two-phase flows has become feasible in the recent years. For industrial applications, the most promising approach seems to be the Eulerian two-fluid model which relies on phase-averaged equations. However, this approach requires many closure relations for which a wide range of sub-models has been developed. Each set of sub-models needs to be validated against small scale experimental data, as we have not yet reached a general model capable of reliably describing different boiling flow regimes. An effort to pave the way towards unified method for testing and validation of two-fluid closure models was made by the NEPTUNE project, where the benchmark test based on publicly available experimental data, has been launched. The first tests will be focused on flow boiling in a simple tube geometry, performed in the DEBORA experimental facility at CEA-Grenoble. DEBORA experiments provide a reliable database on local measurements of boiling phenomena in a simple vertical tube geometry with electrically heated wall. A turbulent boiling flow of Freon R12 or R134a was used to mimic high-pressure conditions, relevant to nuclear applications in pressurised water reactors.

In this work, the boiling flow will be simulated using the Ansys Fluent code. Boiling on a heated wall will be modelled by the heat-partitioning model. The interfacial area or the mean bubble diameter is an essential parameter in the sub-models for momentum, mass and energy transfer between phases. In previous attempts of simulating the DEBORA experiment it was shown that the monodispersed approach is insufficient to properly model the mean bubble diameter. Therefore, a population balance model is used in this work where, bubbles of a certain size are formed on the heated wall and then grow or disintegrate due to evaporation/condensation and coalescence/breakup mechanisms. The aim of this work is to perform a parametric study of different population balance sub-models and their influence on important flow parameters such as gas/liquid volume fraction, liquid velocity and liquid temperature. The calculated results will be compared with the measured data from the DEBORA experiments.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 412

Analysis of bubble breakup sensitivity on fluid properties using Large Eddy Simulations

Jan Kren, Blaž Mikuž

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Two-phase flows play an important role in nuclear power systems during the boiling heat transfer or in the accident management conditions, e.g. during the evaporation of liquid water to steam caused by depressurization in loss of coolant accident (LOCA). An interesting phenomenon in two-phase flows is bubble breakup, which is a challenging process to model in the continuum approximation as the relevant physics takes place at the microscopic scales. Further investigations are needed to control and understand the physics of bubble breakup.
In this paper we present a sensitivity analysis of bubble breakup due to the fluid properties of gas-liquid mixture, such as viscosity and surface tension. We study this phenomenon in a vertical pipe with a diameter of 26 mm and the length of 520 mm. The study is focused on the slug flow regime, particularly a single Taylor bubble in counter-current turbulent flow. Taylor bubble is a long bullet-shaped gas bubble with a diameter almost matching that of the pipe.
The study is performed with Large Eddy Simulation approach in OpenFOAM computer code. We are using the modified interFoam Volume of Fluid (VoF) solver which enables the usage of higher order Runge-Kutta time-integration schemes integrated with PLIC interface reconstruction scheme. Turbulent sub-grid scales are modelled using the Vreman model for eddy viscosity. This setup enables quantitative analysis of the impact of fluid properties on the rate of bubble breakup mechanism.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 413

Design optimization of heat transfer performance in the heads of flow boiling experiment

Anil Kumar Basavaraj, Blaž Mikuž

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Flow boiling is an effective heat transfer mechanism, which is important in many industrial applications including in nuclear power plants. A unique flow boiling experiment has been constructed in Thermal-Hydraulics Experimental Laboratory for Multiphase Applications (THELMA) at Reactor Engineering Division of Jožef Stefan Institute. The experiment consists of a custom-designed heat exchanger, which allows visual observation of the boiling surface. Heat flux at the boiling surface is controlled with the temperature and flow rate of the two fluids involved. The present design provided accurate measurements for low and medium heat flux magnitudes, however, modifications are needed for the flow boiling studies at high heat fluxes.
Numerical simulations provide better understanding of complex devices as well as enable their design optimization. The main objective of the present study is optimization of the heat exchanger, which is used for flow boiling experiments. In particular, previous studies have shown that up to 50% of the total heat transfer in our experiment takes place in the inlet and outlet manifolds, i.e. the heads of the heat exchanger. In order to increase the heat transfer in the test section itself, all heat losses need to be reduced to the minimum, including the heat transfer in the heads of the heat exchanger. For that reason, a computational fluid dynamics (CFD) model has been constructed for the present heat exchanger, which includes conjugate heat transfer in the primary and secondary fluid flow as well as several solid domains that are made of different solid materials. Results revealed the most critical parts of the device with severe heat leakages, which need improvements. Thus, modifications have been proposed in the geometry as well as in the selection of more appropriate material properties. Comparison between the present and the optimized design has shown significantly better heat isolation of the fluids inside the heads, which will hopefully allow experiments at much higher heat fluxes up to the critical heat flux (CHF).

Keywords: Heat transfer, CFD, heat exchanger, heat flux

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 414

Application of enhanced phase-change model for simulation of film boiling around a cylinder

Mihael Boštjan Konèar, Matej Tekavèiè, Mitja Uršiè

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Recent life cycle analyses have identified conventional nuclear power plants (NPP) as one of the cleanest energy sources. In terms of carbon emissions, current operable NPP could be comparable to renewables such as wind and solar. Furthermore, fourth-generation reactors with better fuel efficiency and nuclear safety show immense potential for clean and reliable energy production. One of the technologies of this kind is a sodium-cooled fast reactor (SFR).
Our research focuses on the interaction of melt with sodium during a hypothetical core melt accident in SFR. A rapid and intense heat transfer interaction between the molten core material and the sodium coolant may lead to vapour explosions. At the forefront of our research are the heat and mass transfer mechanisms during vapour explosion in sodium. Experimental investigation with liquid sodium is vastly complex, mainly due to chemical reactivity and opaqueness. Hence numerical studies could deliver valuable insight into the heat and mass transfer mechanisms. On the other hand, vapour explosions are experimentally widely investigated in water. These experiments provide a solid basis for the validation of numerical models.
This study will focus on developing an appropriate numerical model for solving the two-phase flow around a melt particle. Our model will represent a benchmark experiment conducted in the TREPAM (CEA, France) apparatus (Berthound, Int J Thermal Science, 48 (2009), pp. 1728) that mimicked film boiling conditions around a melt fragment. In the TREPAM apparatus, the melt fragment was represented by a heated wire moving at a constant velocity through the pressurised subcooled water.
The two-phase flow will be modelled by the single-fluid approach combined with the volume-of-fluid (VOF) interface tracking method. The simulation applying the Unsteady Reynolds Averaged Navier-Stokes (URANS) approach will be used to resolve the flow. Previous studies have shown that the selection of a phase change model is crucial to solution accuracy. Heat and mass transfer will be studied by the enhanced evaporation-condensation model proposed by Chen (Int. J. Heat Mass Tran., 150 (2020), pp. 119279). The phase-change model will be implemented in ANSYS Fluent using user-defined functions (UDF). High-temperature fluctuations are expected throughout the domain. Therefore, the temperature dependence of fluid properties will be evaluated. The simulation results, in particular the heat fluxfrom the cylinder wall, will be validated with the data from the TREPAM experiment (Berthound, Int J Thermal Science, 48 (2009), pp. 1728).

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 415

Inviscid fluid simulation through Incompressible Schrödinger Flow method: a Finite Element approach

Stefano Riva1, Antonio Cammi2, Carolina Introini3

1Politecnico di Milano Dipartimento di Energia, Via La Masa 34, 20156, Milano, Italy

2Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy

3Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy


Computational fluid dynamics is the standard approach to simulate the behaviour of fluids governed by the Navier-Stokes equations. This problem always involves a suitable treatment of the non-linearity of the advection term in the equations, which is the main bottleneck in performing fast simulations. Moreover, as the Reynolds number increases, the importance of this term becomes larger and larger; for this reason, the Navier-Stokes equations are rarely directly numerically solved, preferring a solution with RANS or LES approaches. These methods model the behaviour of the small scales (totally or partially, respectively), and only the larger scales are directly solved.

In the limit of Reynolds number going to infinite (i.e., viscosity goes to 0), the flow obeys the Euler equations. These equations are still strongly non-linear, and they typically put limitations on the usable time step for stability (a common issue in hyperbolic PDEs). These fluids are referred to as ideal fluids, in which the dissipation given by the viscosity can be neglected.

In 1926, Madelung proposed a hydrodynamical form of quantum mechanics, showing a link between the linear Schrödinger equation and the non-linear Euler ones. In particular, he showed that the latter can be derived from the former, linking the two different physics. Thus, a novel approach to solving complex non-linear PDEs has been proposed, substituting the non-linear Euler equations with the linear one derived by Madelung. The fluid state is now a vector of two complex wavefunctions which satisfy the Schrödinger equation with an incompressibility constraint. This method is called Incompressible Schrödinger Flow, and in literature this problem has been solved using FFT, showing impressive results in the prediction of vortex dynamics.

This work aims at implementing this novel approach in a Finite Element framework so that it is easier to extend it to complex geometries. The results of different simulations will be compared with a classical, state-of-the-art CFD approach. In the future, it would be interesting to investigate the possibility of linking this approach with the temperature equation to include buoyancy effects.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 416


Qingling Cai1, Francesco D Auria2, Jianqiang Shan1

1Xi’an Jiaotong University, West Xianning Road, 28, 710049, Xi’an, China

2University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy


The fast Fourier transform based method (FFTBM) was proposed in the 1990s and is used for accuracy quantification of computer codes. FFTBM provides frequency-based measures for each single TH variables as well as the whole transient calculations. The measurement-prediction discrepancies in the frequency domain are assessed by the average amplitude (AA). An AA close to 0 indicates good agreement between measured and predicted results. AA is dependent to the proper selection of time windows, weighting factors, number of discrete data used. This paper summarized the application of FFTBM from publications in the last 30 years, including the relevant experimental tests, selected parameters and weighting factors, time windows and AAs. It attempts to provide some insights and guidelines for FFTBM application.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 417


Qingling Cai1, Klaus Umminger2, Dominique Bestion3, Francesco D Auria4, Fabio Moretti5, Marco Lanfredini6

1Xi’an Jiaotong University, West Xianning Road, 28, 710049, Xi’an, China

2FRAMATOME, Tour Framatome Cedex 16, 92084 PARIS LA DEFENSE, France

3CEA-GRENOBLE DEN/DTP/SMTH/LMDS, 17 rue des Martyrs, 38054 GRENOBLE CEDEX 9, France

4University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy

5Universita degli Studi di Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione, Largo Lucio Lazzerino 1, 56100 Pisa, Italy

6Nuclear Research Group of San Piero a Grado, San Piero a Grado, 56122 Pisa, Italy


The FONESYS and SILENCE networks are run by some of the leading organizations working in the nuclear sector, and work in a cooperative manner since about a decade having one meeting per year.
The FONESYS members are developers of some of the major system thermal-hydraulic codes adopted worldwide. FONESYS has been created to strengthen the current technology, cooperate and share recent advances, identify and discuss further ways of improvements in system thermal-hydraulic code development and their application especially for licensing purposes and safety analyses.
On the other hand, SILENCE members own and operate important thermal-hydraulic experimental facilities. SILENCE aimed at promoting: cooperation and knowledge transfer; discussion on state-of the art technological issues; revival of interest in significant experimental campaigns; support to organizations and countries embarking in large experimental programs. SILENCE is also promoter of an international workshop on instrumentation and measurement techniques, SWINTH.
In this paper selected key achievements from the networks are presented and some activities proposed to contribute addressing the remaining issues in thermal-hydraulics are summarized.

13.09.2022 15:40 Poster session 1

Thermal-hydraulics – 418

Thermal efficiency of protective cladding layers in liquid sodium-cooled heat sinks containing sharp corners

Nima Fathi, Mahyar Pourghasemi

Texas A&M University, Marine Engineering Tech Department, P.O. Box 1675, Galveston, TX 77553-1675, USA


This work investigates the conjugate heat transfer within Na-cooled heat sinks of different shapes with protective cladding layers on their walls. Liquid metals such as Na and NaK with high thermal conductivity and high boiling temperatures are interesting coolants for applications involving elevated working temperatures and high heat dissipation rates. However, Na and NaK are corrosive liquid metals and react with most commonly used high thermal conductivity solid materials such as copper. On the other hand, high thermal conductivity materials such as copper, silicon, and aluminum are often utilized to fabricate miniature heat sinks in real-world applications. To address this problem, we are modeling flows and heat transfer of Na in copper-based heat sinks of different shapes with stainless steel (SS-316), Inconel 718, and Refractory High Entropy Alloys (RHEAs) cladding layers on their walls. The investigated minichannel heat sinks have sharp corners due to their rectangular, pentagonal, and hexagonal cross-sections. Several different cladding thicknesses of 4.5 mm to 0.5 mm are investigated while the Na inlet Reynolds number varies between 2500-10,000. Obtained local and average Nusselt numbers for cladded and non-cladded heat sinks are compared to evaluate the thermal efficiency of protective cladding layers. Finally, the effect of investigated heat sinks geometric parameters on the thermal efficiency of protective cladding layers is investigated.

13.09.2022 15:40 Poster session 1

Reactor physics – 506

Application of neural networks to neutron data interpolation and evaluation

Sakho Abdoulaye1, Ivan Kodeli2, Pierre-Jacques Dossantos-Uzarralde1

1École Nationale Supérieure d’Informatique pour l’Industrie et l’entreprise, 1, Square de la Résistance, F-91025 Évry, France

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Machine learning is a field that is very much in vogue these days, both in industry and in research. Its possible applications are multiple and varied. In this paper, we will look at one of its potential applications in a sub-field of nuclear physics.

The aim of this work is to determine the efficiency whether the use of neural network methodologys for the evaluation of nuclear data for neutron induced reactions compared to the currently existing methods.
Use of neural networks for the evaluation of neutron data will be discussed for the case of delayed fission yield. This project continues the work started in 2017 on nuclear data and their use in uncertainty and cross-section adjustment analyses. The analyses will be based on experimental data available in the AIEA EXFOR database. TensorFlow, a Machine Learning library created by Google, will be used to deduce correlations between experimental data to derive interpolation laws. The TensorFlow system is a set of tools for running neural networks to solve complex mathematical problems. The applications are based on the Python programming language, while the execution of these applications is done in C++. Knowledge of Python, R and C++ languages will be an asset.

The results obtained will be compared with the predictions of the GEF code which allows analytical calculations of delayed neutron data from the basic parameters and the physical model.
Performance of neural networks will be furthermore tested on other problems such as the evaluation of correlations among various parameters and the Covid propagation.

13.09.2022 15:40 Poster session 1

Reactor physics – 507

On the calculation of adjoint neutron flux in typical PWR for the determination of the neutron flux redistribution factors

Tanja Gorièanec1, Luka Snoj2, Marjan Kromar2

1Institut “Jožef Stefan”, Odsek za reaktorsko fiziko, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


In a typical nuclear power plant, the characteristics of the control rod are determined based on the response of the ex-core neutron detectors. Various methods can be used for this purpose. The rod insertion method was developed in the Reactor Physics Department of the Jožef Stefan Institute and has been successfully used in the Krško Nuclear Power Plant for three decades. During the insertion of the control rod bank, the spatial distribution of the neutron population in the core is significantly changed. Since the detector measures the local neutron flux at the location outside the core, correction functions should be applied to obtain an appropriate control rod worth. To investigate the various effects, a detailed core and ex-core model of a typical PWR NPP was developed for use in the Monte Carlo neutron transport code MCNP. The neutron source required for the calculations was modelled with cylinders on a scale of fuel rod in 24 axial layers. The ex-core neutron detector response can be determined directly or indirectly by multiplying adjoint neutron flux and power distributions. This work focuses on the evaluation of neutron flux redistribution factors using adjoint neutron flux distributions. The reactor core power distributions were determined using a detailed MCNP model of the reactor core, while the adjoint neutron flux distributions were calculated in two ways. First, the MCNP ex-core model was used to determine the source coordinates of neutrons contributing to the detector response, which can be considered an approximate representation of the adjoint neutron flux. Second, distributions of the adjoint neutron flux determined using the ADVANTG code were used. A sensitivity analysis of the ADVANTG data libraries and the geometry representations of the adjoint and power distributions was performed. It was confirmed that the geometry description of the adjoint and power distributions has a noticeable effect on the calculated neutron flux redistribution factors. To verify the results, a comparison was made with the calculated neutron flux redistribution factors obtained by calculating the direct ex-core detector response with the MCNP ex-core model. A pin-wise description of the power and adjoint neutron flux within the core in 24 axial layers gave the best agreement with a deviation from the reference results of up to ~ 2 %.

13.09.2022 15:40 Poster session 1

Reactor physics – 508

Our Experiences with the Benchmark “Rostov-2”

Elina Oberlander, Helmut Glöde, Kai-Martin Haendel

TÜV NORD EnSys GmbH & Co. KG, Am TÜV 1, 30519 Hannover, Germany


The OECD/NEA benchmark „Reactivity compensation of boron dilution by stepwise insertion of control rod cluster into the VVER-1000 core” is based on measurements of neutron physical and thermal-hydraulic behaviour of a water-water energetic reactor VVER-1000. The measurements have been performed at Rostov unit 2 nuclear power plant using 5 different TBC-2M assembly types allowing for an 18-month fuel cycle. The data should be used for the validation of multi-physics codes. For the benchmark Rostov-2 integral (plant) data and local (core) measurement data were provided to the participants for simulations in the course of the benchmark exercises.
As a first step, we determine the cross-sections for the hexagonal fuel assemblies. Therefore, we use two different approaches TRITON/NEWT from the software package SCALE 6.2 (ORNL) and the software package CASMO-5 (SSP) and compare them with each other. In the next step, we simulate the cycle evolution from BOC (Begin Of Cycle) to the initial state of transient of selected core parameters (e. g. boron concentration, radial power peaking factor, volume power peaking factor, core axial offset). Furthermore, we determine the relative assembly power distribution and the axial power distribution of selected assemblies at the initial state of the transient. For the abovementioned simulations of the reactor core behaviour, we use the software SIMULATE-VVER (SSP). During the transient, one control rod group is inserted in the core at constant reactor power and the signals of thermocouples and thermoresistors as well as SPN (Self-Powered Neutron) detectors were recorded and are available for simulations.
We will present our simulation results and compare them to the provided measurement data.

13.09.2022 15:40 Poster session 1

Reactor physics – 509

Modelling gamma calorimetry experiment with JSIR2S code

Klemen Ambrožiè1, Vladimir Raduloviæ1, Hubert Carcreff2, Damien Fourmentel3, Luka Snoj1

1Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France

3CEA Cadarache, IPSN/DRS/SEMAR BAT. 700, 13108 St. Paul Lez Durance, France


Nuclear heating measurements on selected fission and fusion relevant materials were performed in two distinct experimental campaigns at the JSI TRIGA reactor. The experiments were performed using custom, single-cell CEA developed gamma calorimeters containing Eurofer97, 99.95% tungsten, graphite R6650, aluminium Al6063 samples and a reference calorimeter without sample. Due to slow response time of the calorimeter, measurements used were taken roughly 50 min after the reactor start-up. Heating levels of roughly 50 mW/g were obtained for aluminium and Eurofer97, and 112.6 mW/g for tungsten.

During the design phases of the calorimeter, contributions of both neutron and gamma heating were assessed, and was determined that gamma contribution is significantly higher compared to neutrons, except for graphite, where the contributions are approximately even. This means both must be evaluated in detail. Delayed radiation accounts for a roughly 30 % contribution to the total gamma heating. In order to asses this, the JSI developed rigorous two step approach code is used for the calculation of the delayed radiation field, which has been previously validated by faithfully reproducing a variety of dose-rate experiments during reactor operation and after shutdown and even fusion by simulations.

The aim of reproducing the experiments by simulation is to both validate the JSIR2S code for nuclear heating, as well as identifying any possible shortcoming in nuclear data and Monte Carlo particle transport energy deposition techniques. In addition the code is used for evaluation of experimental uncertainties of the measured heating rate. Heating rate profiles are calculated throughout the sample, as well as throughout the calorimeter body in order to compare both obtained nuclear heating power values, as well as to confirm the calorimeter behaviour as a whole.

In this paper, an overview if given on the experiments performed, followed by a detailed explanation of the modelling paradigm with some preliminary calorimeter power profiles.

13.09.2022 15:40 Poster session 1

Reactor physics – 510

Burnup measurements using fuel reactivity worth experiments at the JSI TRIGA Research Reactor

Anze Pungercic1, Alireza Haghighat2, Luka Snoj1

1Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Virginia Tech Northern Virginia Center Mechanical Engineering Department, Nuclear Engineering Program, 7054 Haycock Rd, Falls Church, VA 22043, USA-Virginia


Fuel elements in the reactor core of the TRIGA Mark II research reactor at “Jožef Stefan” Institute have been in use for more than 30 years. The burnup accumulated during this period was studied computationally using the deterministic and stochastic neutron transport codes. The calculations have been validated by comparing measured and calculated excess reactivity changes. In order to validate the individual fuel element burnup, we designed two fuel reactivity worth experiments and conducted them in a 3-day experimental campaign in April 2022. The first experiment was based on the so-called Ravnik’s fuel reactivity worth method [1], in which fuel elements of interest are taken out of the reactor core and inserted into pre-determined reactor core position to measure the difference in excess reactivity of the core, which is directly connected to the difference in fuel burnup. We analysed 7 fuel elements, where reactivity changes were measured using the DMRes system. The measured changes ranged from 30 pcm to 150 pcm. A direct connection with calculated fuel burnup was observed. For the second fuel reactivity worth experiment we designed a new so-called fuel swap method, in which the position of two fuel elements is swapped and change in excess reactivity measured. In this case the difference in reactivity is related to the difference in burnup as well as the positions of the swap. The effect of the position was studied by analysing how the importance function changes due to position and burnup of the fuel elements. With this we were able to determine fuel burnup using the new fuel-swap method and compare it to the established Ravnik’s method. In the full paper, the measurements will be compared to predicted results using the deterministic TRIGLAV code, the Serpent Monte Carlo, and the hybrid RAPID code.

[1] Ravnik, M., et al. “Determination of the burn-up of TRIGA fuel elements by calculation and reactivity experiments”, Kerntechnik 57.5 (1992): 291-295.

13.09.2022 15:40 Poster session 1

Reactor physics – 511

Maximum required excess reactivity due to Xe-135 “poisoning”

Blaž Levpušček1, Gašper Žerovnik2, Luka Snoj2

1Fakulteta za matematiko in fiziko, Jadranska 19, 1111 LJUBLJANA, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


For the stability of the electricity network it is imperative that its production is as close to its consumption at all times. In order to achieve that, the system needs power plants with flexible production capabilities which enables so-called load following operations. Currently, in many countries a large part of load-following is done by fossil fuel powered plants since the capabilities of flexible renewables power sources such as hydro as well as storage capacities are relatively limited. Green transition implies that this part of the load-following will be taken over by other means. One possibility is to use nuclear power plants (NPP), which traditionally operate as base load, i.e. continuously at maximum power. One of the limiting factors that affects NPP production flexibility is the built-up of the strong neutron absorber 135Xe. This effect is called “poisoning” and is most pronounced ~ 10 h after shutdown following a long (> 20 h) operation at maximum power. This limitation is important towards the end of the reactor operation cycle when the excess reactivity of the reactor approaches 0. It is beneficial if the reactor is capable of unrestricted changes of power for as long as possible within the operation cycle, especially in small systems with a few or even only one NPP, such as e.g. Slovenia.

Assuming a point reactor, the required excess reactivity for load-following operation without limitations was estimated for different initial fuel compositions. The coupled neutron transport and fuel depletion calculations were performed using the Serpent code for a typical PWR assembly in 2D geometry with reflective boundary conditions. It is shown that for UO2 fuel, higher initial 235U enrichment results in lower requirements for excess reactivity, thus enabling unlimited load-following operations for a higher fraction of the operation cycle. A similar conclusion can be drawn for MOX fuel with respect to UO2 fuel.

13.09.2022 15:40 Poster session 1

Reactor physics – 513

Moderator Heat Sources in TEPLATOR District Heating SMR

Martin Lovecký1, Tomáš Koøínek2, Jiøí Závorka1, Jana Jiøièková1, Radek Škoda1

1University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic

2Czech Technical University in Prague, Czech Institute of Informatics, Robotics and Cybernatics, Jugoslávských partyzánù 1580/3, 160 00 Prague, Czech Republic


Moderator temperature needs to be maintained below safe values in reactors with separated moderator and coolant volumes. For reactor during full power operation, main heat sources relevant to moderator volume temperature are heat transfer from nuclear fission source and radiation heating caused by various nuclear reactions. These reactions include radiation heating from fission neutrons, secondary photons from (n,g) reactions, fission photons, spent nuclear fuel neutrons, spent nuclear fuel photons and Co-60 photons from activated steel components. Although the radiation heating can represent less intensive heat source, it is a direct source in the moderator volume and it can affect moderator volume temperature more than relatively distant heat sources in the fuel. For reactor during outage before core unloading, heat transfer from nuclear fission is replaced by heat transfer from spent nuclear fuel decay heat. In the paper, radiation heating and its components along with SNF decay heat for TEPLATOR district heating SMR is calculated by MCNP and SCALE codes.

13.09.2022 15:40 Poster session 1

Reactor physics – 515

Calculation and verification of the new neutron absorbers in a well-defined core in LR-0 reactor

Jiøí Závorka, Martin Lovecký, Radek Škoda

University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic


The research is focused on special neutron absorbers designed to improve the optimization of medium-term storage and final storage of spent nuclear fuel. The solution aims to improve nuclear safety, and it is also reasonable from an economic perspective so that the final product can be transferred into practice. The case’s merit is based on fixed neutron absorbers effectively placed within the nuclear fuel assembly. The first theoretical part of the research is devoted to optimizing a suitable material from a neutronic and economic point of view. It is about determining the ratio between the appropriate neutronic properties and the cost of the material and form. The second practical part of the research focuses on the prototype’s production and verification in the research reactor LR-0.

13.09.2022 15:40 Poster session 1

Reactor physics – 516

Neutronic Analysis of Various Fuels for the TEPLATOR HT

Tomáš Peltan, Eva Vilímová, Radek Škoda

University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic


In the light of a new world’s approach focusing on energy decentralisation and decarbonisation, the development of Small Modular Reactors is crucial. The new small reactor TEPLATOR produces low-cost heat for various purposes, such as district heating or process heat. To supply process heat, a high temperature is required. For this reason, a high-temperature version of the TEPLATOR with corresponding fuel is under development. TEPLATOR HT with high output temperature assumes using the organic coolant, which affects the possibility of using contemporary fuels available on the market. This paper focuses on preliminary neutronic analyses that evaluate the coupling of an organic coolant with various available fuel geometries and assesses the feasibility of using certain fuels assuming minimal design changes. All fuel material combinations and geometries were tested in TEPLATOR geometry to choose an appropriate candidate for TEPLATOR HT that can withstand higher operational parameters. Based on the results, it will be decided whether existing fuel can be used for TEPLATOR HT or whether a new fuel type needs to be developed.

13.09.2022 15:40 Poster session 1

Reactor physics – 517

On the effective fuel temperature of the UO2 fuel

Dušan Èaliè1, Marjan Kromar2

1Jožef Stefan Institute, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The fuel temperature is an important parameter in the determining the behaviour of the reactor fuel because the neutron cross sections depend on it. The shape of the temperature distribution in a fuel pellet is generally parabolic when a flat power profile is present. However, due to resonances and self-shielding, the power profile is not flat and changes during irradiation of the pellet. This requires a coupled neutron thermal-hydraulic calculation. Furthermore, in lattice physics calculations, a fuel pellet is usually treated radially as a single region. In this case, the choice of an “effective temperature” that gives the same response as the actual temperature profile is very important. The idea behind choosing the effective temperature is to preserve a parameter of interest, such as the multiplication factor or some other integral value such as isotopic composition, etc. In principle, we can determine specific effective temperatures by averaging the temperature profile with the reaction rates as appropriate weights. However, since the reaction rate profiles of specific reactions (fission on 235U, absorption on 238U, etc.) are different, we obtain different partial effective temperature for each reaction of interest. There are several known techniques for estimating the effective temperature, but they are not general. In this paper, we investigate how best to preserve the masses of important nuclides such as 235U, 238U and 239Pu on the one side and the multiplication factor on the other side by considering the reference results obtained by coupling neutronic and thermal-hydraulic effects by coupling the Monte Carlo code Serpent 2 with the thermal-hydraulic code Finix. The coupled reference results are compared with burnup calculations performed with the stand-alone Serpent 2 runs, with the fuel temperature held constant in the radial direction. The accuracy of some widely used standard techniques is estimated, and some improvements are suggested.

13.09.2022 15:40 Poster session 1

Reactor physics – 518

A Monte Carlo fuel assembly model validation adopting Post Irradiation Experiment dataset

Lorenzo Loi1, Antonio Cammi2, Stefano Lorenzi2

1Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milano, Italy

2Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy


The 3D Monte Carlo code Serpent is currently being validated for Light Water Reactor’s (LWR) fuel cycle simulations. This work chose the Takahama-3 Post Irradiation Experiment (PIE) dataset as a test case. Having key information related to the history of the plant, it was possible to compare the Serpent’s results against more than 35 isotopic species’ concentrations, measured following a destructive analysis of two fuel rods (SF95, SF97) at the end of their irradiation cycle. Nevertheless the presence of systematic sources of uncertainties related to the geometry, the results show a good agreement with the experimental data. Also, it is shown how the prediction capability may be increased up to +8% adopting a realistic temperature mesh for the fuel.

13.09.2022 15:40 Poster session 1

Reactor physics – 519

Dose Rate Assessment around the PCFV Release Line during Severe Accident Conditions in Nuclear Power Plant Krsko

Davor Grgiæ, Paulina Duèkiæ, Vesna Benèik, Siniša Šadek

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia


Passive Containment Filtered Vent (PCFV) was installed in Nuclear Power Plant (NPP) Krsko in 2013 as a part of safety upgrade program. It is intendent for severe accident consequences prevention and mitigation by ensuring the containment integrity. When the pressure in the containment reaches limiting value, the containment atmosphere is released in the environment through the PCFV system exhaust line. But, before released in the environment, the containment atmosphere passes through five aerosol filters in containment and an iodine filter in the auxiliary building to reduce its activity. In this paper, dose rates around the exhaust line of the PCFV system resulting from radioactivity release in case of a severe accident are determined. The assumed severe accident scenario is a beyond design basis station blackout in NPP Krsko, which is simulated by using the MELCOR code. The obtained source term from MELCOR is subsequently used in Monte Carlo calculations. The source is present in the containment, in the iodine filter and in the exhaust pipe. The dose rates around the exhaust pipe are calculated using MCNP6.2 code.

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 603

Strain localization in austenitic stainless steel due to hydrogen concentration

Amirhossein Lame Jouybari1, Samir El Shawish2, Leon Cizelj2

1University of Ljubljana Faculty of Mathematics and Physics , Jadranska 19, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The aging of austenitic stainless steel in the harsh environment of the Light Water Reactor is highly sensitive to Stress Corrosion Cracking. The presence of hydrogen in such steels can change their microstructure and affect the mobility of dislocations, which may result in the deterioration of mechanical properties like embrittlement and strain localization.
This study is concerned with the formation of strain localization in the crystal plasticity finite element model of the austenitic stainless steel polycrystal due to hydrogen concentration. In this framework, polycrystals are generated by Voronoi tessellation topologies with zero crystallographic texture. The hydrogen effect is considered in the decomposition of the deformation gradient into elastic, hydrogen, and plastic parts. A rate-independent form of constitutive equations is derived and implemented numerically in the User MATerial subroutine in Abaqus software. Finally, the effect of hydrogen concentrations is studied in a polycrystalline aggregate in a series of uniaxial tension simulations.

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 604


Olga Ogorodnikova1, Mitja Majerle2, Jakub Cizek3

1Moscow Engineering Physics Institute National Research Nuclear University, “MEPhI”, Kashirskoye shosse 31, 115409 Moscow, Russian Federation

2Nuclear Physics Institute of the CAS, v. v. i., Øež 130, 250 68 Øež, Czech Republic

3Charles University in Prague Faculty of Mathematics and Physics, Prague, Czech Republic


Currently, tungsten and tungsten coatings are the reference materials of the ITER divertor and DEMO reactors and the possibility of using low-activated ferrite-martensitic, RAFM, steels not only as structural materials, but also as the material of the first wall of the fusion reactor is considered. Also, these steels, together with a new generation of RAFM steels with oxide dispersion strengthened by adding Y2O3 nanoparticles, the so-called ODS steels, are considered as promising materials for fast neutron fuel cladding. One of the key ITER and, especially, DEMO issues is radiation-induced damage caused by 14 MeV (in peak) neutron irradiation and its effect on the fuel and helium retention. As a fusion neutron source does not exist yet, to simulate fusion neutron-induced damage in materials, fission neutrons and charged particles are widely used. However, it is not always clear if the mechanisms under the ion irradiation are relevant to lower dose rate and the primary knock-on atom (PKA) spectrum under neutron irradiation. On the other hand, the fusion neutron spectrum is different from that in available fission reactors. In order to simulate the fusion experimental conditions for reliable predictions of radiation damage in fusion reactors, it is necessary to establish the adequacy of the radiation damage produced by different types of irradiation. For this reason, a comparison of radiation-induced defects in metals based on W and Fe produced by high-energy self-ions, protons and neutrons with different spectrum has been performed. Radiation-induced defects have been studied by well-established method of positron-annihilation lifetime-spectroscopy (PALS), transmission electron microscopy (TEM) and nuclear reaction analysis. The study of different distributions of radiation-induced vacancies and vacancy clusters of different sizes created by different types of irradiation using PALS and TEM methods allows us an experimental validation of the value of “displacement per atom” (dpa) when comparing different types of irradiation. We found a formation of the larger size of the defects with lower density in the case of irradiation with high-energy neutrons from the p(35 MeV)-Be source compared to fission neutron- and proton- irradiations. It is shown that fission neutrons do not appear to be a good surrogate for simulating radiation damage caused by thermonuclear neutrons. Fast neutrons from p-Be source or other accelerator sources can be a good surrogate to simulate radiation damage caused by fusion neutrons. Energetic protons can be a surrogate to simulate fusion neutron damage in certain materials over a certain temperature range. The new experimental data together with data available from the literature are compared with the dpa theory, including molecular dynamic simulations. Second, He/dpa ratios in different neutron facilities have been compared. We show that He/dpa ratios in the facilities with the hard energy spectra (fusion like) p(35 MeV)-Be source and DEMO are one-two orders larger than in the fission ones LVR-15, HFIR and BOR60. Methods to obtain the best approach to modelling fusion neutron damage and to bridging the gap between theory prediction of primary defect formation and long-term damage, including gaseous and solid transmutation products, as well as thermal effects (including the temperature gradient in the normal operation regime and during ELMs) are discussed taking into account the uncertainties.

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 605

Heat Capacity of PuO2 at High Temperature: a comparison of interatomic potentials

Rolando Calabrese

ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy


A new generation of fast breeder reactors (FBRs) is under development with the objective of making nuclear energy more sustainable. Most promising reactor designs are loaded, at least during their early phase of deployment, with UO2-PuO2 mixed oxide fuel (MOX). Concentrations of plutonium dioxide that are foreseen for FBR range up to 30 mol%. This highlights the need for a sound and deep knowledge of the thermophysical properties of PuO2. Evaluations on PuO2 heat capacity are usually carried out by using the Neumann-Kopp rule confirming previous statement. Heat capacity is important for evaluation of the thermal conductivity and performance under transient conditions. However, measurements on the heat capacity of plutonium dioxide are scarce or even lacking at high temperature. Numerical methodologies such as MD calculations have been employed to overcome the difficulties encountered in experimental measurements. Besides numerical also theoretical models have been applied as valuable tools for interpretation of enthalpy measurements. Nevertheless, due to the mentioned lack of experimental measurements issues such as the existence of the Bredig transition and the formation of defects at high temperatures are still debated in nuclear fuel research. Excess enthalpy seen in measurements of actinides oxides has been explained by means of either electronic disorder or anion disorder. In the case of plutonium dioxide, a common consensus has been reached on the hypothesis that anion disorder leads to a significant increase of heat capacity at high temperature. Konings and Beneš have developed a model that accounts for this phenomenon. Their correlation has been often included in models of heat capacity and employed for recommendations. However, in the high temperature region MD calculations showed an underestimation of model predictions that was not compensated by the presence of a peak of heat capacity that has been interpreted as the Bredig transition. Based on these observations, this paper presents MD evaluations on the heat capacity of PuO2 at high temperature that are mostly focused on the formation energy of oxygen Frenkel pair and its correlation with the model proposed by Konings and Beneš. An interatomic potential published in the open literature and developed in compliance with the experimental thermal expansion of PuO2 is taken as reference. Coefficients of this model have been modified aiming at implementing a value of formation energy of oxygen Frenkel pair (OFP) that could be consistent with values in the open literature. This paper presents a comparison of MD calculations that have been obtained by applying the reference and modified interatomic potential. The discussion is mainly focused on results of heat capacity at high temperature. Besides this predictions on other relevant quantities such lattice constants, melting temperature and elastic constants are also presented.

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 606

Crystallographic phase transition of zirconium alloys: new models for the TRANSURANUS code

Rolando Calabrese

ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy


TThe TRANSURANUS fuel performance code is featured by a clearly defined mechanical and mathematical structure which has permitted since its beginning a continuous development and an extension of the domains of application. In this view, the TRANSURANUS team has devoted significant efforts to make the code applicable for loss-of-coolant accident (LOCA) calculations. In parallel, besides standard Zircaloy-2 and Zircaloy-4, cladding material correlations for E110 that is used in VVER western-type reactor and, more recently, for the M5 alloy of Framatome have been introduced in the code based on information in the open literature. Conditions occurring during LOCA and Reactivity-Initiated Accidents (RIA) may induce a crystallographic phase transition of zirconium alloys with a consequent degradation of mechanical performance of the cladding. Advancements in modelling of Zircaloy-4 and M5 phase transition have been published recently. Based on these findings, our paper presents revised correlations having the objective of improving accuracy of beta fractional volume predictions especially at high values of heat rate or introducing the effect of quantities that are not accounted for in the original model, such as hydrogen concentration for M5. Presented activities have been carried out in the frame of the Reduction of Radiological Consequences of design basis and design extension Accidents project (R2CA) of EURATOM.

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 607

Development of a 3D-RPV Finite Element Model for Pressurized Thermal Shock Analyses

Oriol Costa Garrido1, Nejc Kromar2, Andrej Prošek3, Leon Cizelj3

1Institut “Jožef Stefan”, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Univerza v Ljublani,Fakulteta za strojništvo, Aškerèeva cesta 6, SI-1000 Ljubljana, Slovenia

3Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The reactor pressure vessel (RPV) is an indispensable component in nuclear power plants and its structural integrity must be assured under all possible events. The limiting event for the long-term operation (LTO) of the RPV is the pressurized thermal shock (PTS). A PTS event typically follows a loss-of-coolant accident (LOCA), or other emergency scenarios where the subsequent injection of cold water from the emergency core cooling system into the hot RPV may induce high thermal stresses in the RPV wall. The RPV wall material undergoes neutron embrittlement after several years of operation, with the subsequent hardening and loss of fracture toughness. PTS analyses are thus needed to assure that a potentially existing flaw in the RPV wall will not initiate and propagate rapidly in a brittle-fracture manner during LOCA scenarios.
This paper presents the development of a full three-dimensional (3D) finite element model of a RPV with four cooling loops. The goal of the paper is to generate the necessary model meshes to accurately analyze the temperatures and stresses developing in the RPV during a small-break LOCA (SB-LOCA). To that end, several meshes are developed with different element densities. The inner surface of the RPV is assumed to be subjected to time-dependent and uniformly-distributed fluid temperature, heat-transfer coefficient and pressure, representative of an SB-LOCA transient. These same loads are used in a parallel analysis with the FAVOR code, which assumes a 1D model (in the through-thickness direction) of the RPV wall. The temperatures and stresses obtained with the developed meshes and the FAVOR code are then compared. The outcomes of the comparison include the selected mesh for accurate results and reasonable computational resources to perform the analyses, as well as the impact on the results from the use of 1D and 3D RPV wall models. This work has been performed in partial fulfillment of the European project APAL (Advanced PTS Analysis for LTO).

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 608

Direct conditioning of molten salt arising from the thermal treatment of solid organic waste

Anna Cerna, Vojtìch Galek, Petr Pražák, Jan Hadrava

Research centre Rez, Hlavni 130, 250 68 Husinec-Øež, Czech Republic


During the operation and decommissioning of nuclear facilities, organic radioactive waste is generated. This includes both solid (spent ion-exchange resins) and liquid (scintillation cocktails, oils, organic solvents) wastes. Therefore, it is essential to focus on the possibilities for processing, reduction, and disposal, not just due to its radioactivity content but also due to its chemical composition.
It is expected that not all organic radioactive wastes will be suitable to direct conditioning due to their high volume and instability in the expected conditions in the final repositories. Thermal treatment offers a potential route to process this type of waste where Molten Salt Oxidation (MSO) was identified as one of the possible pathways for radioactive organic waste treatment. In the MSO process, the organic waste is dosed, together with oxidising medium, under the surface of the molten salt, where flameless oxidation takes place. The non-combustible inorganic substances, such as heavy metals or radionuclides, are trapped in the molten salt, which can be further processed.
The study aimed to determine the possibility of direct conditioning of resulting molten salt, which, arises as the secondary waste after the combustion of spent ion exchange resins in the geopolymer matrix. After initial tests, a geopolymer of the commercial name LK was chosen as the most suitable choice. The series of experiments were then performed with 5, 10, 15, 20, 25, 30, 35 and 40 %wt. of spent MSO salt added to the matrix. Adding more than 40%wt alkali salt into the matrix wasn’t possible as the mixture could no longer be thoroughly stirred. The samples were cured in different conditions such as in mold, air dry, and in the dryer for 24 hours at 65 °C. Mechanical strength and XRD composition analysis were performed on the prepared samples. The results have shown an increased mechanical strength after adding 20 %wt. or more alkali salt.

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 609

Thick protective Fe and Mo coatings for PbLi coolant environments prepared by RF-ICP and cold spray

Jan Cizek1, Jakub Klecka2, Lukáš Babka1

1Institute of Plasma Physics of the Czech Academy of Sciences, Za Slovakou 3, 18200 Prague, Czech Republic

2Institute of Plasma Physics, Czech Academy of Science, Za Slovankou 3, 18200 Prague, Czech Republic


Progress in developing generation IV nuclear fission reactors and fusion systems entails many challenges to overcome. One of the potential concepts foresees the possible use of liquid metal-based cooling media, where the most promising candidates are heavy liquid metals such as lead-lithium eutectic (PbLi). This material excels in thermal conductivity, which is crucial for rapid and efficient heat transfer needed in the cooling systems of the fission reactors. Using such medium requires proper protection of the structural materials to prevent degradation processes, e.g., corrosion. In Pb or PbBi environments, such protection is typically achieved by maintaining low oxygen levels in the liquid metals, triggering a formation of surface passivation layers. In PbLi, this principle cannot be used due to the high affinity of Li to O2. Here, one of the solutions could be a deposition of protective, long-term stable coatings of pure metals onto the surfaces. Two deposition techniques were used in this study, radio frequency inductively-coupled plasma spray (RF-ICP) and cold spray (CS). Both methods can deposit thick coatings with good adherence, and, importantly, without oxidation. Due to their favorable properties, Fe and Mo were deposited onto two structural steels (Eurofer, ODS Eurofer). Several powder types were tested and the spray processes were optimized. The coated steels were then tested in stagnant, liquid PbLi environment at 600 °C for 500 and 1000 hours. The most promising results were achieved using atomized Fe powders. After a detailed study of the results, it can be stated that these coatings were successfully able to prevent the degradation of the structural material.

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 610

Comparing different approaches to intergranular-stress modelling

Timon Mede, Samir El Shawish

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The knowledge of intergranular normal stresses is essential for predicting the process of intergranular stress-corrosion cracking, which is the most frequent cause for failure of polycrystalline components in corrosive environments under mechanical loading. To obtain the local grain-boundary stress basically means solving the constitutive equations for all the grains in the aggregate. While this can be done numerically, for instance by relying on the finite-element method, such an approach is both expensive and time-consuming, not to mention impractical, since it requires the information on the exact configuration of all the grains in the aggregate, including their shapes, crystallographic orientations and possible defects (voids, inclusions, …). On the other hand, such computation is a bit of an overkill, since what we are in fact after, is estimating the probability that a macroscopic crack would develop in a certain industrial component when specific external loading is applied to it and for that the precise microstructure of each piece is not relevant. Additional problem is that not all grain boundaries can withstand the same amount of stress before they crack, i.e., they have different grain-boundary strength. In principle that strength depends on the complete neighbourhood of the grain boundary. But since its effect gradually diminishes with distance, the two adjacent grains are the most relevant and in first approximation we can neglect all others. Then each grain boundary is defined by its orientation with respect to external stress (2 degrees of freedom) and the type it belongs to (specified by 5 parameters for a chosen material). The idea is that grain boundaries of the same type also have the same strength, while those classified into different types can in principle differ.
The simplest modelling approach is to treat the material as ideally elastic and solve the Hooke’s law for such pair of grains embedded in a homogeneous and isotropic matrix material. This is called the bicrystal model. Its solution requires the same number of boundary conditions as there are constitutive equations, in this case 12 for all the stress-tensor components in both grains. While the ‘’internal’’ boundary conditions on the grain boundary are straightforward, additional 6 ‘’external’’ conditions are needed to relate the strain of a bicrystal pair to external stress. If these were known, the model could be solved exactly, but unfortunately that would again require solving the equations for all the grains in the aggregate. The simplest assumption then is that the bicrystal deforms as the bulk material of the same size under external loading would. With this the model can be solved, although not analytically (except in some special cases) due to the mixed nature of boundary conditions – some apply to strains while others to stresses. However, it is easy to understand that for soft grain boundaries the resulting stress magnitude is too small (since in reality the grains should deform more than the bulk) while for stiff grain boundaries the effect is the opposite. To loosen that constraint, we introduce some ‘’buffer’’ grains, i.e., we embed the grain pair in linear chains and demand that each chain as a whole deforms like bulk material. To solve this new model analytically, we do not invoke the complete set of internal boundary conditions, in particular we neglect the strain compatibility across the grain boundaries, which in turn simplifies the treatment of shear-stress components. In this paper we introduce both models and investigate how different assumptions used in both (boundary conditions, use of buffer grains, …) affect their results, among others the normal-stress distributions and their first two statistical moments for various grain boundary types, dependence on the twist angle, relevance of effective stiffness, …

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 611

Neutron irradiation effects investigation on the silicon nitride (?-Si3N4) nanoparticles using EPR spectroscopy

Elchin M. M. Huseynov1, Sahil P. Valiyev2

1Institute of Radiation Problems of Azerbaijan National Academy of Sciences, B.Vahabzade 9, AZ1143, Baku, Azerbaijan

2National Nuclear Research Center, Inshaatchilar pr. 4, AZ 1073, Azerbaijan


Neutron irradiation effects investigation on the silicon nitride (?-Si3N4) nanoparticle using EPR spectroscopy
Elchin M. Huseynov1-2, Sahil P. Valiyev2
1Institute of Radiation Problems of Azerbaijan National Academy of Sciences, AZ 1143, B.Vahabzade 9, Baku, Azerbaijan
2Department of Nanotechnology and Radiation Material Science, National Nuclear Research Center, AZ 1073, Inshaatchilar pr. 4, Baku, Azerbaijan

Such as in other nanomaterials silicon nitride has characteristic properties at nano scale. Nanomaterials at high temperature, ionizing environments and under mechanical influence represent distinctive properties. Moreover, nanomaterials usually are very sensitive and changing physical properties of nanoparticles are extremely difficult and actual issues. However, by the neutron flux it is possible to influence properties of some silicon based nanomaterials. In general approach neutron irradiation effects on silicon based and other class nanomaterials were studied in some extent [1-5]. During the neutron irradiation Si3N4 nanoparticles naturally will produce different type defects. Investigation of nature and identification of these defects is vital and actual issue. Electron Spin Resonance (ESR) or EPR spectroscopy method is one the leading and unique methods used to study defect cases inside the materials and investigate of nature of some defects. In this work creation and nature of defect cases in Si3N4 nanoparticles was comparatively studied by EPR spectroscopy before and after neutron irradiation.
EPR spectroscopic analysis was performed before and after neutron irradiation at magnetic field values of 0.05 – 0.55 T (500 – 5500 Gauss). The 3460G-3580G area corresponding to Si-based paramagnetic centers is discussed in detail. After neutron irradiation some annihilated EPR signal explained. The formation mechanisms of Si-Si, Si3?Si* and Si?N3-based paramagnetic centers and the effect of neutron-induced transformation on them has been studied. Before irradiation, signals from various 4 centers were observed in Si3N4 nanoparticles. As a result of the neutron flux influence, the two signals recombined and disappeared. 31P isotopes and other neutron effects cause the loss of other g3 and g4 centers observed in Si3N4 nanoparticles. It has been shown that the g1 signal is a narrow signal and corresponds to the hole capture in the Si-Si bonds. It is also known that similar spectra can be characterized by localized electrons in the Si?N3 state. It has been found that the relatively broad g2 spectrum can be characterized by dangling bond such as Si3?Si* silicon-based.

1. Elchin M. Huseynov, Tural G. Naghiyev “Various thermal parameters investigation of 3C-SiC nanoparticles at the different heating rates” Applied Physics A 128, 115, 2022
2. Elchin M. Huseynov “Thermal stability and heat flux investigation of neutron-irradiated nanocrystalline silicon carbide (3C-SiC) using DSC spectroscopy” Ceramics International 46/5, 5645-5648, 2020
3. Ravan Mehdiyeva, Elchin Huseynov “Effects of Neutron Irradiation on the Current–Voltage Characteristics of SiO2 Nanoparticles” Silicon 10/4, 1369–1373, 2018
4. Elchin M. Huseynov, Adil A. Garibov, Sahil P. Valiyev “EPR study of silicon nitride (Si3N4) nanoparticles exposed to neutron irradiation” Radiation Physics and Chemistry 195, 110087, 2022
5. Elchin Huseynov, Adil Garibov “Effects of neutron flux on the temperature dependencies of permittivity of 3C-SiC nanoparticles” Silicon 9/5, 753–759, 2017

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 612

Properties and prospects of bulk W-based composites with DBTT at 200 °C

Petra Jenuš1, Aljaž Ivekoviè2, Anže Abram1, Andrei Galatanu3, Magdalena Galatanu4, Elena Tejado5, Jose Ygnacio Pastor5, Marius Wirtz6, Gerald Pintsuk6, Saša Novak7

1Jožef Stefan Institute, Department of Nanostructured Materials K-7, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Institut “Jožef Stefan”, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

3National Institute for Laser, Plasma and Radiation Physics, P.O. Box MG36, Magurele-Bucharest, Romania

4National Institute of Materials Physics, Strada Atomi?tilor 405A, Mãgurele 077125, Romania

5Universidad Politécnica de Madrid Dpto. de Ciencia de Materiales-CIME, Calle Ramiro de Maeztu, 7, 28040 Madrid, Spain

6Institute for Energy and Climate Research Forschungszentrum Juelich GmbH, Juelich, Leo-Brandt-Straße, 52428 Jülich, Germany

7Jožef Stefan Institute, Department for nanostructured materials, Jožef Stefan International Postgraduate School, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Tungsten is considered the material of choice for the divertor application of fusion power plants due to its intrinsic thermo-physical properties. However, its bulk DBTT temperature and the reduction of its mechanical properties at elevated temperatures are governing research in a quest for its improvement. This work aims to improve the tungsten’s properties to sustain plasma-facing conditions in the divertor. Among the available options, we selected the reinforcement of tungsten with carbide nanoparticles (W2C), wherein the reinforcement should not chemically react with the matrix.
W-based composite was formed in-situ during the thermal treatment of powder mixture consisting of W and WC particles (4 at % of carbon in the form of WC nanoparticles, sample denoted as W-4WC) with a field assisted sintering technique (FAST). In addition to the microstructural and phase analysis, thermo-mechanical properties at room and elevated temperature and high-heat-flux tests were carried out. Thermo-mechanical properties measured up to 1000 °C revealed the materials’ DBTT is at 200 °C. With the satisfying thermal conductivity, which does not drop below 100 W/m K at elevated temperatures (up to 1000 °C), and promising HHFT behaviour in PSI-2, this composite makes an interesting alternative to the pure tungsten.

Acknowledgement: This work has been carried out within the framework of the EUROfusion Consortium, funded by the European Union via the Euratom Research and Training Programme (Grant Agreement No 101052200 — EUROfusion). Views and opinions expressed are however those of the author(s) only and do not necessarily reflect those of the European Union or the European Commission. Neither the European Union nor the European Commission can be held responsible for them. Slovenian Research Agency is acknowledged for funding the research programs P2-0087 and P2-0405 , CEMM, JSI for the use of EM.

12.09.2022 18:20

Materials in nuclear technology – 613

Towards More Ductile Refractory High-Entropy Alloys at Room Temperature

Sal Rodriguez1, Eric Lang1, Sharpe Rob1, Erin Barrick1,Darryn Fleming1, Andrew Kustas1, Levi Van Bastian1, Graham Monroe1,Nima Fathi2

1Sandia National Laboratories, 1515 Eubank, SE, P.O. Box 5800, Albuquerque, NM 87123-1379, USA

2Texas A&M University, Marine Engineering Tech Department, P.O. Box 1675, Galveston, TX 77553-1675, USA


The 2010 advent of refractory high-entropy alloys (RHEAs) for high-temperature aerospace applications resulted in a flurry of thousands of research papers, thereby stemming the manufacture of hundreds of different RHEA combinations. Some of the RHEA combinations have shown remarkable properties that exceed the performance of Inconel 718, such as high-strength at elevated temperature, corrosion resistance, erosion resistance, self-healing, and creep resistance. However, it is noted in metallurgy that materials with high strength tend to have reduced ductility, and the converse is true. As noted over the past few years, the number of high-strength RHEAs with ductile properties at room temperature (RT) is rather scarce, with much less than 1% of RHEAs achieving this metric; most RHEAs are notoriously brittle at room temperature, though fortunately, not all. Ductility is a key driver for RHEA manufacturing and commercialization purposes, i.e., widespread marketability, because ductility is intricately associated with the necessary machinability of industrial components. Certainly, high-strength ductile (HSD) RHEAs are of much interest not just to the aerospace industry, but also to other industries, such as energy and transportation.
Here, a search of the literature associated with HSD RHEAs at RT was conducted, fully realizing that this field continues to grow at an accelerated pace, so it is nearly impossible to find all such references. In any case, a vital RHEA researcher noted in 2018 that just HfNbTaTiZr and a few of its derived, hybrid combinations fulfilled such metric. Herein, four years later, about a dozen such combinations were identified in the literature. A table of such potentially-HSD RHEAs was compiled, and were recently manufactured by our team via spark plasma sintering (SPS) and laser engineered net shaping (LENS). The RHEAs were characterized and tested experimentally to determine various key properties associated with machinability and strength, including tensile ductility, hardness, yield strength, as well as their relative capability to withstand drilling and lathing operations. Based on these observations, a synthesis of their elemental combinations, material properties, and machinability provided various insights regarding promising HSD RHEA compositions and pathways for improving ductility at RT.
This paper describes objective technical results and analysis. Any subjective views or opinions that might be expressed in the paper do not necessarily represent the views of the U.S. Department of Energy or the United States Government. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.

13.09.2022 15:40 Poster session 1

Materials in nuclear technology – 614

Mitigating environmentally-assisted cracking in light water reactor environments through optimisation of surface condition – results of the MEACTOS collaborative project

A. Sáez Maderuelo1, T. Austin2, R.-W. Bosch3, M. G. Burke4, M. Grimm5, M. Herbst5, A. Hojna6, T. Kosec7, A. Legat7, A. Maurotto8, P. J. Meadows9, R. Novotny2, V. N. Olaru10, T. Pasutto11, F. J. Perosanz Lopez1, Z. Que12, S. Ritter13, V. Román Flórez14, F. Scenini4, A. Toivonen12, A. Treichel13, M. Vankeerberghen3, B. Zajec7, M. Zimina6

1CIEMAT, Madrid/Spain

2JRC, Petten/The Netherlands

3SCK CEN, Mol/Belgium

4University of Manchester, MPC, Manchester/UK

5Framatome GmbH, Erlangen/Germany

6CVR, Řež/Czech Republic

7ZAG, Ljubljana/Slovenia

8NAMRC, Sheffield/UK

9Jacobs, Warrington/UK

10RATEN ICN, Pitesti/Romania

11EdF, Moret sur Loing/France

12VTT, Espoo/Finland

13PSI, Villigen/Switzerland

14ENSA, Maliaño/Spain


The goal of the Horizon 2020 MEACTOS (Mitigating Environmentally-Assisted Cracking Through Optimisation of Surface Condition) collaborative project is to improve the safety and reliability of Gen II and III nuclear power plants by improving the resistance of critical locations, including welds, to environmentally-assisted cracking (EAC). The main objective was to determine how different surface machining procedures could be used to mitigate EAC in some typical light water reactor structural materials and environments. The surface of austenitic stainless steel (SS) type 316L (cold-worked) and Ni-based weld metal Alloy 182 specimens have been machined in different ways (ground: RS, face milling: STI, face milling in supercritical CO 2: SAM1, SAM1 + minimum quantity lubrication: SAM2, shot peening: SP). The EAC initiation susceptibility of these specimens was first screened by accelerated constant extension rate tensile (CERT) tests under simulated boiling (BWR) and pressurized water reactor (PWR) conditions. Tapered tensile specimens were used as the main advantage of using such a geometry is, that in a single test a stress gradient is obtained through the gauge length, and therefore a stress threshold for crack initiation can be determined by electron microscope investigation of a single specimen after the test. The results of the screening tests were then verified also by constant load (CL) experiments in same environment. Scatter in the results of the accelerated EAC initiation testing limited the trends that could reliably be observed, whereby only minor or even no clear improvements of surface grinding (RS) or advanced machining (SAM) compared to the standard industrial face milling were revealed. While the results from the constant load tests confirmed the stress thresholds for EAC initiation in most cases, a fully conclusive picture of the EAC initiation behaviour for all materials and conditions has yet to emerge. In the current poster, the summary of most important results and conclusions from this five-year collaborative project is presented.

13.09.2022 15:40 Poster session 1

Radioactive waste – 704

Corrosion study of carbon steels in contact with alkaline pore water saturated cement-bentonite grout or cement paste at deep geological disposal conditions

Bojan Zajec1, Petra Moènik1, Andraž Legat1, Jules Goethals2, Charles Witterbroodt3, Valéry Detilleux4, Tadeja Kosec1

1Zavod za gradbeništvo Slovenije, Dimièeva 12, 1000 LJUBLJANA, Slovenia

2IMT Atlantique, 4 Rue Alfred Kastler, 44300 Nantes, France

3Institut de Radioprotection et de Sureté Nucléaire, 31, avenue de la Divison Leclerc, 92260 Fontenay Aux Roses, France

4Bel V, Rue Walcourtstraat 148, 1070 Brussels, Belgium


Several national concepts for the geological disposal of nuclear waste are based on cement-bentonite grout or cement paste as a backfill material in clayey host rock. After certain exchange time, the chemical composition of the solution saturating the grout material will correspond to a mixture of native grout pore water and natural argillaceous rock pore water. This solution may not have sufficiently high pH to warrant the passivity of carbon steel overpack, particularly due to the dilution by argillaceous pore water.
Our study will investigate two types of carbon steel associated to different chemical composition and also microstructure. The outcome of several types of experiments, all carried out in anoxic conditions, will be presented:
Electrochemical characterization of both steels in synthetic groundwater (pH ? 11 and pH ? 13.4) at several temperatures (room temperature, 37°C and 80°C). This comprises open circuit potential measurement, linear polarization resistance, electrical impedance measurement and potentiodynamic polarization scan.
Periodic in-situ electrochemical characterization (open circuit potential measurement and linear polarization resistance) of both metals while immersed in the synthetic groundwater being in contact with the solid cement-bentonite grout, at high temperature. It is expected that these long-term measurements would reveal the temporal evolution of the possible passivity breakdown.
The true extent of the corrosion damage can only be revealed after the exposure unless the electrical resistance sensors are being used for corrosion monitoring. Our in-house developed electrical resistance sensors for corrosion measurement will be immersed in the synthetic groundwater being in contact with the solid cement-bentonite grout, at several temperatures. This will help to monitor the intensity of corrosion damage (average depth of corrosion damage) during the long term exposure.

13.09.2022 15:40 Poster session 1

Radioactive waste – 705

Thermal Modeling of SNF Behaviour During Dry Storage

Luis E. Herranz1, Francisco Feria1, Jaime Penalva2, Michela Angelucci3, Sandro Paci3

1As. CIEMAT, Av. Complutense, 40, 28040 Madrid, Spain

2IDOM ENGINEERING AND CONSULTING S.A.U., Avenida Monasterio de El Escorial, 4, 28049 Madrid, Spain

3University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy


The safe storage of Spent Nuclear Fuel (SNF) in dry conditions previous to its final disposal in a deep geological repository should meet a number of regulatory requirements. One of them is keeping its structural integrity. In the case of in-cask dry storage, a number of phenomena might jeopardize the fulfilment of this requirement, from cladding creep to embrittlement due to the radial precipitation of hydrogen absorbed by the cladding during the irradiation in the reactor. These mechanisms are strongly affected by the cladding thermal state, as it is reflected in the criteria to be met to guarantee cladding integrity: temperature limits are set at 673 K and 843 K for normal and off-normal conditions, respectively. In other words, an effective SNF cooling should be ensured, which emphasizes the importance of a suitable modelling of dry cask thermal performance.
The present contribution gives an overview of the thermal modelling of dry casks from the experience gained by CIEMAT in the course of research. A description of fundamentals will be succinctly given, while necessary support for approximations made when using 3D CFD will be defended. Particular attention is given to the challenges related to model verification and validation and what alternative simpler methods, relying on the MELCOR lumped parameter code and on a Python algorithm based on the heat transfer theory, may bring up in terms of phenomenological insights and/or licensing requirements. At the end of the paper a discussion on what is needed for further development and what drawback might limit such developments will be discussed.

13.09.2022 15:40 Poster session 1

Radioactive waste – 706

Estimation of Dose Rates around Dry Storage Building during Campaign One Loading in Nuclear Power Plant Krsko

Paulina Duèkiæ, Davor Grgiæ, Mario Matijeviæ, Radomir Jeèmenica

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia


In this paper neutron and gamma (fuel gamma, neutron induced gamma and hardware activation gamma) dose rates were estimated around the Dry Storage Building (DSB) in Nuclear Power Plant (NPP) Krsko during Campaign one loading using MCNP6.2 and ADVANTG3.03. The Campaign one consists of 16 HI-STORM storage casks filled according to NPP Krsko specific fuel loading plan. The characteristics of the spent fuel were based on real operating history and their source in terms of neutron and gamma intensity and spectrum are calculated using ORIGEN-S module from SCALE6.2.4. The annual dose at the closest site boundary and dose rates at the DSB walls are compared with the regulatory limits of 0.5 mSv and 3 µSv/hr, respectively.

13.09.2022 15:40 Poster session 1

Radioactive waste – 707

Development of the Vrbina LILW Repository Design

Boštjan Duhovnik

IBE, d.d., Hajdrihova 4, 1000 LJUBLJANA, Slovenia


Planning activities for the LILW repository project began shortly after the start of regular operation of Krško NPP. Generic conceptual design solutions of the LILW repository were prepared in 1987 for two alternatives: shallow ground disposal and tunnel-type disposal. Design solutions were adopted and placed at potential locations, identified in the second step of the LILW repository site selection process in Slovenia, which took place in years 1990 – 1992.
After the interruption of the site selection process and the cancelation of the results, the analysis of all possible disposal concepts was systematically carried out. In addition, the Krško NPP underwent changes in conditioning and packaging of waste, which significantly affected the concept of disposal. Considering these facts, revised generic conceptual solutions were developed for disposal into surface disposal cells and near-surface tunnel-type disposal in 1999.
The generic solutions were technologically updated and expanded with the disposal alternatives within the scope of the process of placing the LILW repository in the space in 2004. For the design solution – disposal into below-ground silos, which was evaluated as the most suitable disposal alternative for the Vrbina site, the conceptual design documentation and the preliminary design documentation were made; followed by the optimization of the preliminary design, which was completed in 2011.
Considering optimized design solutions and thorough field research at the Vrbina site, the elaboration of the Design documentation for obtaining a construction permit started in 2014, focusing on the open issues of the structural stability of disposal silo, especially from the point of view of the seismicity and nuclear safety. At the same time, the process of developing, testing and certifying of the disposal container was carried out. Adopted design solutions were elaborated in detail in the Design documentation for implementation of construction, which was finished at the end of 2021.

13.09.2022 15:40 Poster session 1

Radioactive waste – 708

Numerical simulation of cemented RLOW simulants for packaging system

Rosa Lo Frano, Salvatore Angelo Cancemi

University of Pisa, Department of Civil and Industrial Engineering, Largo Lucio Lazzarino 2, 56124 Pisa, Italy


The selection of immobilization waste technologies based on cementation should demonstrate that the so obtained matrices for packaging systems are reliable, durable and stable. Cementation of radioactive liquid organic wastes (RLOW) is a difficult technological task because of complex chemical composition, and relatively high activity of wastes streams to immobilize. A proper characterization of the cemented matrices is therefore felt necessary.
The aim of this study is to investigate numerically, by means of finite element (FE) modelling, the thermo-mechanical behaviour of cementitious material or RLOW simulant. The FE numerical model is adopted to benchmark several RLOW simulants composition and correlate/compare the structural properties. Ageing effects are also investigated.
Results are compared to experimental data. They indicate that the thermal conductivity monotonically decreases as the temperature increases. The compressive strength confirmed to be dependent on w/c ratio and to suffer irradiation damage; e.g. it reduces as porosity increases

13.09.2022 15:40 Poster session 1

Radioactive waste – 709

Characterization of copper corrosion in bentonite slurry using coupled multi-electrode arrays

Miha Hren, Tadeja Kosec, Bojan Zajec, Andraž Legat

Zavod za gradbeništvo Slovenije, Dimièeva 12, 1000 LJUBLJANA, Slovenia


Long-term exposure of copper to bentonite is known to cause localized corrosion attack on the copper surface, if the bentonite slurry is saturated with groundwater, is under oxic conditions and contains chloride ions. The Canadian concept of spent nuclear fuel storage includes steel canisters, which are covered by copper as protective material. If the copper gets damaged, Galvanic corrosion between steel and copper may occur, which could result in either formation of a localized corrosion or repassivation.

In the present study, copper corrosion will be characterized by means of coupled multi-electrode arrays (CMEAs). Two scenarios will be examined: copper corrosion as the surface is mechanically scratched, and galvanic corrosion between copper and steel. The CMEA technique will provide us with information on how cathodic and anodic locations develop over time, specifically when different events (e.g., scratching, galvanic corrosion) occur. The experiment will take place in bentonite with simulated saline groundwater and in oxic conditions. These are the same conditions as present in Canadian spent nuclear fuel storage containers. In addition to the CMEA technique, microCT will also be used to calculate the damaged volume of copper and steel electrodes. After the exposure, spectroscopic techniques will be applied to determine the type of corrosion products and validate CMEA and microCT results.

The goal of the study is to determine the magnitude of both localized and average corrosion rates, the type of copper and steel corrosion, and how corrosion rates develop over space and time.

13.09.2022 15:40 Poster session 1

Nuclear energy and society – 805


Jure Jazbinšek, Garsia Kosinac, Melita Lenošek Kavčič, Tomaž Žagar

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia


Paper presents case study based on 10 years of experience and interactions with visitors in GEN Interactive visitor center “Svet energije” in Krško / Slovenia. Misconception or myths about nuclear energy such as: “Nuclear powerplants have cooling towers which are basically giant chimneys that emit smoke”, or “Serious possibility of nuclear (bomb-like) explosion of NPP” are some of common ideas that average visitors express during their visit. During multiple interactions with visitors and numerous public discussions, where such misleading ideas were expressed, we developed optimised responses to such situations. It is very important, to handle explanation of misconception in a gentle, non-dismissive tone and try to debunk a nuclear myth with particular attention to prevent possible embarrassment or ridicule a person that expressed this misconception.

13.09.2022 15:40 Poster session 1

Nuclear energy and society – 806

Safeguards-related Activities in Slovenia since 2000 – Some Outcomes in a Nutshell and a Quick Look forward

Janez Èešarek

Slovenian Nuclear Safety Administration, Litostrojska ulica 54, 1000 Ljubljana, Slovenia


This article would like to address the Slovenian experience with the safeguards activities, i.e. those activities by which the international inspections from the International Atomic Energy Agency (IAEA) and the European Commission (Euratom) can verify that Slovenia fulfils its international commitments not to use nuclear programmes for nuclear-weapons purposes. The Slovenian state system of accountancy and control will be presented together with some major changes after the Slovenian membership in the European Union (EU) and the implication of the “Additional Protocol” to safeguards agreement.

Beside the known international treaties and agreements, the Slovenian legislation does provide sound basis for safeguards-related obligations, commitments and confidence-building measures so as to assure that all nuclear material and activities are intended for peaceful purposes only. Both the Ionising Radiation Protection and Nuclear Safety Act as well as the Resolution on Nuclear and Radiation Safety in the Republic of Slovenia (for the period 2013-2023) are two important pillars.

The Slovenian Nuclear Safety Administration (SNSA) has been in a vivid interaction with domestic holders (and users) of different kinds of nuclear material. In addition, the interaction of the Member States of EU, i.e. their nuclear regulators, with the European Commission (Euratom) is also salient, not to miss out the regulators’ own engagements and new(er) topics and interfaces – which have emerged recently.

The inspections of IAEA and the European Commission (Euratom) have been a regularity (also) in Slovenia and their number has been fairly stable regardless of the introduction of so called integrated safeguards years ago. The SNSA’s staff has participated during nearly all international inspections; this proactive approach has not been important only for a bunch of minor holders of nuclear material but also for nuclear facilities – in particular when inspections are conducted based upon “Additional protocol”.

Over the years, and as it has been a case in most countries worldwide, smaller quantities of nuclear material (by definition) have been found. It will be briefly touched upon why such historical sources – or sudden orphan sources in scrap – are the issue – and how to establish adequate accountancy and control of such nuclear items (“batches” in safeguards terminology).

Covid-19 has had certain, fortunately quite limited influence on safeguards-related activities (e.g. inspections or “paperwork”) and SNSA has pursued a proactive approach vis-a-vis international organisations and domestic holders of nuclear material.

Last but not least, the Commission Regulation (Euratom) No 302/2005 on the application of Euratom safeguards has been in force for a number of years, and some insights will be addressed to unveil necessity for possible future amendments of this relevant regulation.

13.09.2022 15:40 Poster session 1

Nuclear energy and society – 807

Background of Values Used for Exclusion, Exemption and Clearance of Practices and Sources from Regulatory Control

Matjaž Koželj, Vesna Slapar Borišek

Jožef Stefan Institute, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


During the licensing or registration process and also during the use of radioactive sources officials, applicants, and later licensees or registrants, meet with some kind of the classification method when a practice or radioactive source is categorised. The purpose of this approach is to simplify the legal decision process and to optimise requirements regarding required documentation and requests for safety measures and emergency preparedness.

Authorities use different categorisations for different purposes: the first and basic one is the decision of whether some practice or material should be controlled. Some practices and sources are excluded from regulatory control, and other exemption criteria are used for the decision on whether regulatory control is required. And when some source should be released from regulatory control, the clearance level is used.

For practically all radionuclides all these levels are conveniently listed in relevant regulations and available to all users. But the real background of the values is not evident to the majority of people involved in the authorisation process and use of sources. Therefore, we have decided to review and explain the logic behind exclusion from regulatory control and the background of criteria and scenarios that were used to develop and calculate the exemption and clearance levels. We hope that these explanations could be of practical value for professionals dealing with the assessment of exposures of members of the public and professionals from sources of ionising radiation.

13.09.2022 15:40 Poster session 1

Nuclear energy and society – 808

Public Opinion about Nuclear Energy – Year 2022 Poll

Radko Isteniè1, Igor Jenèiè2

1Retired from JSI, Jamova 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The Information Centre which is part of the Nuclear Training Centre at the Jožef Stefan Institute informs the visitors about nuclear power and nuclear technology, about radioactivity, about Krško Nuclear Power Plant and about energy in general.
Our main target population are the schoolchildren from the last grades of elementary school and from high school (ages 13-18) with their teachers. In the last decade we had close to 8000 visitors per year, but in the years 2020 and 2021 we had a significantly lower number due to limitations imposed by covid-19 pandemic. We expect a partial recovery in 2022. The visitors can choose between live lectures on nuclear technologies (fission and fusion), a lecture about use of radiation in medicine, industry and science and a lecture on stable isotopes. A general lecture about energy and an energy workshop is also available. The visit includes a demonstration of radioactivity, a tour of our permanent exhibition and a virtual tour of the TRIGA research reactor.
Since 1993 we monitor the opinion trends by polling some 1000 youngsters. There are 10 questions in the poll and they remain unchanged for several years. This enables us to follow the trends in the basic knowledge of energy issues among youngsters and their attitude towards nuclear energy.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 907

International Cyber Security Exercise at Nuclear facilities KiVA2022

Samo Tomažiè, Saša Kuhar, Metka Tomažiè

Slovenian Nuclear Safety Administration, Litostrojska ulica 54, 1000 Ljubljana, Slovenia


The ongoing global situation has once again proven that the nuclear sector needs to be highly protected from physical attacks as well as cyber-attacks. To accomplish this, many assurance activities have to be implemented. One example of such activities is cyber security exercises. The value of exercises is even higher if all relevant national and international stakeholders are involved and if the interface between all domains in the nuclear sector is covered. Cyber security exercises provide an opportunity to verify and validate organizational capability to protect, detect, respond, and recover from cyber-attack. An example of this kind of exercise was KiVA2022, which was organized and conducted by the Slovenian Nuclear Safety Administration in cooperation with the International Atomic Energy Agency and the Austrian Institute of Technology. KiVA2022 highlights this by demonstrating contemporary approaches to nuclear safety, security, and emergency preparedness.
Taking place over three days, KiVA2022 was divided into two parts, namely the VIP event and the exercise. The first day was intended for the VIP guest, which was attended by numerous high representatives of organizations involved in ensuring cyber security in the nuclear sector in Slovenia. On this day, the developers of the exercise presented objectives, scenario, equipment, and efforts put into the development of KiVA2022.
In the following two days, key cyber security stakeholders of the Slovenian nuclear sector attended the exercise. They were divided into four groups: operator, regulator, technical support organization and computer equipment supplier. The operator was represented by all three nuclear facilities in Slovenia (Nuclear Power Plant Krško, Central Radioactive Waste Storage and TRIGA Research Reactor). The regulator was represented by Slovenian Nuclear Safety Administration, Government Information Security Office, and Ministry of the Interior. Finally, the technical support organization was represented by SI-CERT, SIGOV-CERT, and the Ministry of Defense. Computer equipment supplier was represented by Iskra d.o.o. and Agitavit Solutions d.o.o., which are both actual suppliers in the Slovenian nuclear sector. The exercise was also observed by many international observers from Austria, Argentina, Romania, Switzerland, the United Arab Emirates, and the United States of America.
The exercise was conducted in a fictional country of Anshar in a fictional nuclear facility Asherah. Scenario of the exercise involved insider threat, demonstrations, blended cyber/physical attack and cybersecurity compromise.
For the exercise, mock-ups of systems that are representative of those found in nuclear facilities were developed. The first was a physical protection system, that supports multi-factor authentication and surveillance functions. The second one was a hardware-in-the-loop system that includes representative OT equipment, virtual IT and OT systems, and a simulator of the Asherah NPP. In addition to IT and OT systems, injects involved videos, webpages, contracts, and penetration testing equipment like Rubber Ducky, LAN Turtle and Bash Bunny.
The exercise has once again proven Slovenia’s preparedness to respond to security incidents, including those caused by cyber-attacks, as well as strong connections between national and international stakeholders. This type of the exercise should be repeated at a regular interval recognizing cyber security as one of the most important part in nuclear sector in the future.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 908

Qualified digital protection devices

Martin Nano

Framatome GmbH, Paul-Gossen-Str. 100, 91052 Erlangen, Germany


Electrical systems have to be protected against faults to prevent loss of human life, to limit damages and to allow continued operation of the parts of the system which are not directly impacted by the fault itself (selectivity). These protection devices measure electrical parameters like current, voltage, frequency, temperature and arc protection and compare them with certain limits.
Multifunctional digital protection devices are state of the art in the industry for decades. They provide several advantages compared to electromechanical and (non-software based) electronic devices:
-No drift of setting values
-Fault recording capabilities
-Better sensitivity of protection
-Faster reaction times
-Combination of several protection functions into one single device (less cost, less space)
-Communication capabilities (between devices and for monitoring)
-Self-monitoring capabilities (inputs, power supply, memory, processor)
-Possibility of modular systems, which can be adapted to the requirements
However, these digital devices are based on microprocessor technology and use software/firmware.
In the nuclear domain, use of software requires special qualification measures and processes especially if it is used for safety classified equipment and systems.
2.Failure modes
Electrical protection is capable to interrupt even actions of the reactor protection system and has therefore a high safety importance.
There are basically two different failure modes of electrical protective devices:
-“Under function”, i.e. no trip in case of fault
-“Over function”, i.e. trip in case of an external trigger without fault
In case of nuclear application, there are always redundant systems available in parallel, i.e. if there is an electrical fault not detected/switched off by electrical protection only one system is lost and the redundant systems continue to operate. Therefore “under function” is not critical for safety classified redundant systems, as several identical electrical faults in redundant systems have not to be postulated (provided separation and isolation rules have been followed).
However, “over function” in case of external triggering events (e.g. transients in voltage, frequency or current) could lead to common mode failures occurring in redundant systems leading as worst case to total loss of the safety system. Therefore, this failure mode is the critical one in the nuclear domain.
Besides mistakes in the setting of the protection relays and the implementation of the protection function, software failures could lead to such common mode failures.
Proper software qualification according to nuclear standards is therefore mandatory in case of application in safety systems.
3. Software qualification
Unfortunately, electrical protective devices are developed in large quantities for industrial purposes and usually nuclear standards are not considered during its development. The challenge is now how to qualify such an industrial “black box” product following nuclear standards.
The main nuclear safety standard for software qualification is IEC 60880 for cat. A applications. For cat. B and C it is IEC 62138. For industrial safety applications, the most common standard applied is IEC 61508 (SIL categories).
The approach selected is now a delta comparison of the pre-existing software in the protection relay with the requirements of IEC 60880 (IEC62138) and IEC 61508 SIL3 (SIL2). As the software was not developed according to these standards, some deviations are identified.
To cover these deviations, the use of diverse protection devices is unavoidable. If one type of relay fails it has to be ensured that at least one defense lines remains.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 909

Seasonal and daily operation of nuclear based district heating system with varying energy demand

Jan Škarohlíd1, Radek Škoda2, Tomáš Koøínek3

1Czech Technical University, Zikova 1903/4, 166 36 Prague 6, Czech Republic

2University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic

3Czech Technical University in Prague, Czech Institute of Informatics, Robotics and Cybernatics, Jugoslávských partyzánù 1580/3, 160 00 Prague, Czech Republic


Operation of nuclear based district household heating is connected with several challenges. These challenges mostly reflect two aspects. The first aspect is reflecting daily based changes of heat consumption with morning and evening demand peaks. Second aspect is connected with huge difference in summer and winter demand, where summer demand (mostly for district hot water production) is usually about 10 % of winter nominal consumption. Nuclear reactors usually prefer stable operation on nominal power, without fast changes and are not designed for rapid power changes. In this study we would like to introduce a concept of a nuclear district heating station, for real situation with realistic heat consumption profile. We will address changing heat consumption on daily and seasonal basis.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 910

Optimization of Sizing and Operation of a Nuclear District Heating System Using Teplator and Heat Storages

Hussein Abdulkareem Saleh Abushamah1, David Masata2, Jana Jiøièková2, Radek Škoda2

1University of West Bohemia, Faculty of Electrical Engineering, Univerzitni 8, 301 00 Pilsen, Czech Republic

2University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic


Expansion of the nuclear energy-based district heating systems can effectively contribute to the elimination of the CO2 emissions, reduce the dependency on the fossil fuels, supply the heating demand with higher efficiency, and are eligible to be competitive compared with the individual heating systems. Consequently, the interest in using the heat-only small modular reactors for district heating applications is growing, and some concepts such as Teplator are recently proposed. However, a flexible operation and fast load following are required, while generally, nuclear plants are designed to be operated at nominal capacity, and boosting flexibility at nuclear plants is technically complicated. Therefore, using heat storage or auxiliary boilers could be effective for load following and peak shaving, which also may reduce the total costs by decreasing the required capacities of the nuclear plant and heat transmission system. In this study, the optimization of the design and operation of a nuclear district heating system using heat storage and auxiliary boilers is formulated. The capacities and hourly-based operation of the heat generation units and the charging and discharging schedule of the thermal storage are optimized. The technical constraints, such as maximum charging/discharging power rates of the heat storage, or the power ramp rate of the heat generation units, are included in the model.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 911

Simulation of load following operation with a PWR reactor

Dušan Èaliè1, Luka Snoj2, Gašper Žerovnik2

1Jožef Stefan Institute, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


As the share of intermittent renewable energy sources grows, so does the need for on-demand response electric power generation. In order to asses limitations of existing typical PWRs to load follow operation in combination with intermitten renewable energy sources, such as wind and solar, we first developed load follow scenarios. In load-following scenario, the power is changed several times a day. From a neutronic point of view, several changes occur in the core: The fuel and moderator temperature changes, the xenon concentration and distribution are modified, the power distribution is skewed axially, etc. These changes must be appropriately balanced to keep both the core critical and the power distribution acceptable. The traditional approach in pressurised water reactors (PWRs) is to compensate for the reactivity changes due to the power variations by adjusting the soluble boron concentration and moving a limited number of control rod banks. Then we used these scenarios to simulate the load following operation of a typical PWR using the JSI developed LOADF package.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 912

Modernization of the Fuel Assembly Register Software

Sebastian Pleško, Slavko Slaviè, Bojan Žefran, Marjan Kromar, Luka Snoj

Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The Fuel Assembly Register (FAR) software, developed in 1994 at the “Jožef Stefan” Institute, covers all aspects of nuclear material accounting in a nuclear facility. FAR was originally developed for the Krško NPP, but due to the generality of the programming and the modularity of the structure, it can be easily applied to any other nuclear facility. The main purpose of the package is to simplify the control and updating of fuel properties during the NPP lifetime. Automatic data processing reduces the possibility of errors and allows easier implementation of a quality assurance programme.
The software was first created in 1994 as a DOS code in the Clipper programming language using dBase for data storage and later received numerous upgrades, including a rewrite for Windows in Alaska. Now, in 2022, it has been rewritten again in an intranet form, with the interface running in a web browser. The backend of the software is written in the PHP language and uses the SQLite database, while the front end uses HTML/JS/CSS/SVG. It can work on a single computer locally or as an intranet application on the local network. It retains essentially all of the previous functionalities (except for the deprecated ones), but places more emphasis on the graphical interpretation of data, which is welcomed for the better human-computer interaction. Because computers are orders of magnitude faster today than they were decades ago, some tasks, such as calculating isotopic data and decay heat, have been greatly simplified and are now automatic with increased accuracy.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 913

Dynamic Electrical Simulation Model of NPP Krsko

Luka Zidariè1, Aleksandar Momirovski1, Milos Maksic1, Dragotin Vehovar2

1Elektroinštitut “Milan Vidmar”, Hajdrihova 2, P.P. 285, 1000 Ljubljana, Slovenia

2Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia


Republic of Slovenia is a process of simultaneous green energy development and digitalization. A very reasonable step to facilitate and accelerate such transition, which is required by the EU Commission Regulations for System Operations and Requirements for Generation units as well as by the European Network of Transmission System Operators for Electricity (ENTSO-E), is a creation of a simulation tool i.e., a specified digital twin of the real electric power system. In the digital environment such tool enables simulations of electromechanical transients in the electric power system and provides detailed observability of the power system dynamics and stability – a challenge which becomes even more important because of the implementation of the Renewable Energy Sources (RES) in parallel to the pillars of the power system stable operation i.e., the Nuclear Power Plants (NPP). In that respect a digital twin of the existing NPP Krško is proposed.

The paper presents an electrical simulation model of nuclear power plant (NPP) Krško. The model consists of a stationary and dynamic RMS (Root Mean Square) model of NPP Krško. The simulation models include the modelling of generator, turbines with the associated turbine governor, excitation system with the associated automatic voltage regulator (AVR), Power System Stabilizer (PSS), under-excitation, over-excitation and stator current limiters, generator step-up transformers, and supplementary consumption. The simulation model is parameterized based on technical documentation of the NPP building blocks and comparison of measured and simulated quantities. The simulation models are made in the DIgSILENT PowerFactory 2021 SP5 software package. Each simulation model is also subjected to simulation tests with the aim of proving the robustness of the model.

The digital twin of the NPP Krško shows the importance in the wider power system operations. NPP Krško significantly contributes to the stability and reliability of the Slovenian electrical power system. NPP Krško crucially preserves the power system inertia and thus improve its electromechanical disturbance immunity and essentially enables the development of the unpredictable RES.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 914

Introduction of KHNP O&M Technology

Sangwoo Kim

Korea Hydro & Nuclear Power, 655, Bulguk-ro, Gyeongju-si, Gyeongsangbuk-do, Korea , 38120, South Korea


KHNP has continued construction and operation of nuclear power plants and not only developed a new Korean-designed advanced reactor, but is also developing an up-to-date O&M technology in cooperation with the supply chain. KHNP would like to introduce proven technologies in Mechanic(BHI) and I&C(REALGAIN) fields which has abundant experiences in supplying equipment to nuclear power plant not only domestic but also oversees NPPs. The BHI is specialized in manufacturing pressurized equipment such as Feedwater Heater and Condenser with certificate for European Market. And REALGAIN is the one of the leading company participating in refurbishment for I&C equipment such as Radiation Monitoring system and DRCS indication system in Korea and BNPP.

13.09.2022 15:40 Poster session 1

Nuclear power plant operation – 915

Resampling a Detailed Reactor Power History With the Cycle Power History Script

Vid Merljak

Krško Nuclear Power Plant, Vrbina 12, 8270 Krško, Slovenia


For various applications a reactor power history dataset might be needed. One example is the determination (calculation) of isotopic composition of spent nuclear fuel. Sometimes we need data with great detail (i.e. raw input data), sometimes with fewer detail. This is why the Cycle Power History Python script was developed. It allows the user to specify the number of intervals with which the input data is to be approximated. Furthermore, a few different methods are available for automatic plateau- and step-change detection. Among these, a relatively simple interval merging – based on distance from average of the two neighbouring values – has proven to be both accurate and robust. This article explains the methodology and provides some showcase results.

13.09.2022 15:40 Poster session 1

Fusion – 1004

Influence of Ti/Te ratio on the SOL transport properties

Jernej Kovaèiè1, Stefan Costea2, Tomaz Gyergyek1

1Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Institut “Jožef Stefan”, Odsek za reaktorsko fiziko, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The scrape-off-layer (SOL) of a tokamak works as an exhaust of the core plasma, where the energy is being produced. Since this layer is in direct contact with the solid surfaces, its properties need to be carefully engineered and operated to achieve the least possible amount of plasma-facing-component (PFC) damage at the highest possible reactor yield. Much is still lacking in the understanding of SOL physics, since here collisional mean free paths range from very short to very long, depending on the charged particle and neutrals density. Our research is trying to expand the knowledge on the transport of particles and energy in the parallel direction of the magnetic field using a kinetic particle-in-cell (PIC) code.
In our work we have studied the influence of temperature ratio Ti/Te between the ions and the electrons on the properties of the plasma filaments that are the main mechanism of cross-field transport. This ratio seems to be very important in the SOL transport and is often present in operational regimes due to different fold-off lengths for electron and ion temperatures. We were able to simulate several cases with different temperature ratios and absolute temperatures of particle species in the source region. The results of the simulations were temporal developments of fluid observables as well as kinetic properties such as energy distribution functions of particles which showed that Ti/Te indeed has significant impact on the SOL transport properties. Interestingly, the high absolute temperatures in combination with high temperatures seem to have a big influence on the transport of energy to the terminating surfaces when injected filaments go beyond a certain size. We were able to deduce some general rules on the scaling of the transport depending on the Ti/Te ratio.

13.09.2022 15:40 Poster session 1

Fusion – 1006

W-Cu composites for advanced divertor concepts

Aljaž Ivekoviè

Institut “Jožef Stefan”, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Current design of the water-cooled divertor in fusion reactors is based on a joint structure consisting of a tungsten (W) armour and a copper-based heat sink. However, dissimilar thermo-mechanical properties (melting temperature, thermal conductivity, coefficient of thermal expansion) of the two constituent materials impose severe constraints on fabrication and use of such material combination. High heat flux (10-20 MW/m2) imposed on the divertor components combined with cyclic thermal shock events, lead to the accumulation of plastic strains and fracture at the bonded interface between the two materials.
The goal of the conducted research was to develop a W-Cu heat sink material with enhanced reliability, thermal conductivity and control of the coefficient of thermal expansion (CTE). In order to overcome the challenges of component design and enable greater freedom in terms of composition, W-Cu composites were produced by a combination of additive manufacturing and liquid melt infiltration. W lattice structures were manufactured using laser powder bed fusion (LPBF) followed by infiltration with molten Cu. A series of samples was produced with variation in Cu content from 5-60 vol. % and evaluated in terms of thermal, electrical and mechanical properties. Additionally, in order to verify the feasibility of fabrication and to minimize the effects of the high mismatch in CTE (CTECu ? 4CTEW), a functionally graded W-Cu composite was produced. Final material exhibited comparable thermo-mechanical properties to the conventionally manufactured W-Cu materials, however with an expanded composition range (and shape complexity). On the other hand thermal expansion coefficient was lower across the entire composition range (4.5-5.8 ×10-6 K-1) which can be ascribed to W connectivity throughout the material. Combination of additive manufacturing in liquid metal infiltration offers great potential in component design and optimisation of the divertor performance.

13.09.2022 15:40 Poster session 1

Fusion – 1007

Tungsten particle-reinforced copper composites as advanced heat sink materials for divertor application

Diana Knyzhnykova

Institut Jožef Stefan, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Divertor is a vital component of the fusion device, designed to extract heat and plasma impurities and to protect the surrounding walls from thermal and neutron loads. Therefore, it is subject to the highest heat loads in the reactor. According to the current design, the DEMO divertor consists of serial array of rectangular units, tungsten monoblocks, connected to a copper alloy (CuCrZr) cooling tube running through the central region of the monoblocks. Tungsten serves as a functional plasma-facing armour material, whereas, the copper alloy tube acts as a structural heat sink to remove the heat from the first wall. Maintenance of structural integrity under high-heat-flux (HHF) fatigue loads is a critical concern for assuring the reliable HHF performance of a plasma-facing divertor target component. Next to the failure of plasma facing W due to plastic low cycle fatigue (LCF) cracks, failure at the W-Cu interface as a result of stress accumulation and/or neutron embrittlement of the Cu interphase is a major concern.
To mitigate this risk of interphase failure, a functionally graded transition between the heat sink and the armour material is proposed. W lattice structures were manufactured using field assisted sintering technique (FAST) followed by infiltration of Cu melt into a porous preforms, resulting in a W-Cu composite material. A series of W samples with variation in porosity was produced by variation of initial W particle size. During molten Cu infiltration the effects of temperature and atmosphere on the infiltration process were evaluated. Final W-Cu composites were characterised in terms of density and microstructure.

13.09.2022 15:40 Poster session 1

Fusion – 1008

STOK – a tool for parametric modelling of simple tokamaks

Anže Gabrijel, Aljaž Èufar

Jožef Stefan Institute, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


In fusion and fission reactor design and analyses, CAD (computer-aided design) models are usually the staple when it comes to geometry preparation. However, the preparation of these models is often a time-consuming process requiring a significant amount of CAD designer time. Furthermore, this process is usually cyclical, where a design evolves through several iterations based on the results from various analyses. This begs the question, can such models be created using less user time and are such computer-generated models practical for use in neutronic simulations?

This paper focuses on describing the methods behind and usage of a parametric modelling tool called STOK, a Python-based program that allows parametric generation of simple tokamaks with rectangular cross-sections. These models are generated using sets of parameters, for instance, the vacuum vessel requires the user to define five parameters and when all the required parameters are defined a model is created using these parameters as constraints. Furthermore, intersections and other foreseen geometrical conflicts between components are automatically resolved, e.g., ports into the tokamak automatically introduce suitable openings into the shape of the vacuum vessel. Once the model is complete its components are exported in .step and .stl formats which can be directly used in neutronic simulations or can be converted using translators into a suitable format for use in simulations, e.g., through conversion into constructive solid geometry (CSG) commonly used in Monte Carlo codes for nuclear analyses. With these two methods we can, firstly, perform cross-comparisons between different neutron transport codes like MCNP and Serpent for a set of significantly different geometries in a relatively automated fashion and secondly, verify and benchmark different methods of geometry import into these codes, e.g., defining geometry with CSG in Serpent and importing it directly as .stl files to quantify the differences in both results and simulation efficiency.

13.09.2022 15:40 Poster session 1

Fusion – 1009

Towards automation of fusion reactor design optimization – neutronic optimization in simple parametric models

Aljaž Èufar, Anže Gabrijel

Jožef Stefan Institute, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Fusion reactor design and optimization can be a time-consuming task. One of the challenges in expediting this process is the time needed to perform analyses that test the performance of proposed designs. In neutronics analyses often the most time-consuming part is the preparation of the geometry. Typically, both the design of systems and their integration in the reactor can undergo multiple iterations and the preparation of new geometric models can significantly slow down the rate of iteration. As a consequence, the design optimization can be rather slow and only a limited number of possible design solutions can be explored.

In some cases, the solution to speed up the design modifications is parametric modelling. While the preparation of parametric models from scratch is typically more time consuming than preparing a single model in nonparametric fashion, the use of the same foundations for a number of models can prove to be more efficient. This is especially effective when simplified models or models with significant amount of repeated structures defined by a small number of parameters can be used. The use of tools such as PARAMAK or STOK which further simplify the generation of such models through the use of pre-prepared parametric reactor components can additionally speed up the process of setting up the groundwork for preparation of large number of different models.

If such models can be easily used in analyses then they can also be used in automatic design optimization schemes. In this paper we investigate different schemes for automatic design optimization of neutronic performance in some relatively simple fusion-relevant models. It is expected that experience with optimization algorithms, selection of initial designs, and a choice of fitness function used in such optimizations will to some extent translate into optimization of more complex models useful for more realistic studies and design optimizations.

13.09.2022 15:40 Poster session 1

Fusion – 1010

Analysis of Large Helium Ingress into the DEMO Cryostat Using MELCOR for Fusion

Rok Krpan, Janez Kokalj, Mitja Uršiè, Matjaž Leskovar, Boštjan Konèar

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


The purpose of our research is to analyse the loss of cryostat vacuum due to helium (He) ingress. The initiating event considers a failure of cryogenic lines used for cooling of superconducting magnet coils, resulting in spillage of large amount of He into the cryostat atmosphere, which is normally kept at very low pressure (vacuum).

MELCOR for Fusion code, which is based on the MELCOR code version 1.8.6, will be used to perform the analysis. MELCOR is an integral engineering-level code that includes several models for describing the physics during the considered accident phenomena. The main fusion related modifications of the MELCOR code, relevant for this analysis, include cryogenic He or air as a primary fluid, and modifications for water freezing and ice layer formation. Input modelling is based on the “control volume” approach that describes the fusion reactor system fluid volumes and solid structures. The MELCOR model for the analysis of He ingress into the DEMO cryostat includes control volumes representing the cryostat, the space between the vacuum vessel (VV) and the vacuum vessel thermal shields (VVTS), the interspace between VVTS and the cryostat thermal shields (CTS) that encloses the magnets, and also the interspace between the cryostat and the bioshield. Each control volume is connected by several flow paths representing the lower, equatorial and upper ports, which enable heat transfer by natural convection. Heat transfer due to conduction and radiation between heat structures that include cryostat walls, VV walls, VVTS, CTS, all magnet coils and the bioshield wall is also modelled.

Since the DEMO cryostat is designed as a thin-walled structure, the cryostat boundary fails at the pressure of 1.05 bar with opening of a rupture disk. A parametric study will be performed to analyse the pressurization of the previously vacuumed atmosphere inside the cryostat. The total He inventory in the DEMO magnet coils is estimated to 20 tons, but different amount of gas can fill the atmosphere inside the cryostat, due to different response times of the cut-off valves. Furthermore, also several different break dimensions and different heat transfer coefficients will be considered. On the other hand, for the purposes of this study, vacuum vessel and superconductive magnets will be kept at a constant temperature.

13.09.2022 15:40 Poster session 1

Fusion – 1011

Simulation of Natural Convection in DEMO Cryostat During Helium Ingress Accident

Matija Založnik, Rok Krpan, Martin Draksler, Matej Tekavèiè, Mitja Uršiè

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


One of the hypothetical accident scenarios in DEMO fusion reactor considers ingress of helium into the cryostat due to the break of the cryogenic cooling line, which is used to cool the superconducting magnets. The atmosphere inside the cryostat is normally kept at very low pressure (vacuum) to minimize heat transfer between cold structures (superconductive magnets and thermal shields) and warm structures (vacuum vessel and cryostat).

The helium at cryogenic temperatures is released into the atmosphere inside the cryostat, and since the cryostat is surrounded on the outside by air at room temperature, the heating of the released gas inside is the main cause of the pressure increase. Since the cryostat is designed as a thin-walled vacuum vessel, it could rupture due to overpressurization.

The steady-state simulations of the natural convection is performed using ANSYS CFX computational fluid dynamics code. A numerical model including a 22.5° sector of the interspaces between the vacuum vessel and the cryostat inner wall was developed. The computational model contains vacuum vessel (VV) with port structures, superconducting magnet coils, vacuum vessel thermal shield (VVTS), cryostat thermal shield (CTS) and the cryostat. Three compartments are considered in the model: the interspace between the VV and VVTS, the interspace between the VVTS and CTS and the volume between the CTS and the cryostat. The aim of this study will be to determine temperature distribution, heat fluxes, and heat transfer coefficient at different pressures and temperatures. Namely, these parameters, especially the heat transfer coefficient, are needed to properly assess the heating of the helium gas and the pressure increase inside the cryostat.

13.09.2022 15:40 Poster session 1

Fusion – 1012

Development of a Thermohydraulic Model for DTT PFU

Patrik Tarfila1, Boštjan Konèar1, Oriol Costa Garrido2, Giacomo Dose3, Francesco Giorgetti4, Selanna Roccella4

1Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Institut “Jožef Stefan”, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

3University of Rome, Piazzale Aldo Moro, 5, 00185 Roma RM, Italy

4Department of Nuclear and Radiological Safety National Atomic Energy Agency, Konwaliowa 7A, 03194 WARSAW, Poland


The development of a thermohydraulic model of the divertor plasma-facing units (PFUs) for the Divertor Tokamak Test (DTT) facility [1] will be presented. A computational fluids dynamics (CFD) simulation at a steady state regime will be performed, representing an operating condition relevant to the magnetic equilibrium in which DTT will operate (single null (SN) scenario). Input data for the simulation include inlet thermohydraulic parameters, heat loads (obtained from the physics simulation) and the geometric model of the DTT PFU. A coupled solid-fluid CFD model of the PFU will be developed. The mesh sensitivity analysis will be performed on a smaller model to select the final mesh for the full-scale CFD simulation. The expected results of the simulation represent relevant thermohydraulic parameters such as pressure drop, temperature difference of the coolant between the inlet and outlet and temperature distribution in PFU solid structures. Numerical simulations will be performed with the ANSYS CFD code.

[1] G. M. Polli, “Divertor Tokamak Testing Facility (DTT): A Test Facility for the Fusion Power Plant”, (2021) Offshore Mediterranean Conference and Exhibition, 978-88946678-0-6

13.09.2022 15:40 Poster session 1

Fusion – 1013

Kinetic Effects in ITER Scrape-off Layer

Ivona Vasileska, Leon Kos

University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerèeva cesta 6, 1000 Ljubljana, Slovenia


The divertor targets in tokamaks are constantly bombarded with high-energy neutral and charged particles and such violent events can pose a serious threat to the long-time resistance of the divertor materials. The wall erosion, caused by the bombardment, releases impurities, that migrate towards the bulk plasma and due to the effects, the plasma state is deteriorated. In order to keep the limits of wall erosion, it is important to estimate the plasma characteristics in the Scrape-off Layer (SOL) i.e the region outside the last closed magnetic surface (separatrix). However, the transient heat loads such as ELMs (Edge-Localized modes) occur in tokamak edge during H-mode confinement lead to a significant loss of stored plasma energy. Once the ELM-driven plasma pulse has crossed the magnetic separatrix, it travels mainly parallel to the magnetic field lines and ends up hitting the divertor plate.

This contribution describes the first results of efforts to address this issue for ITER simulations under high-performance conditions using the 1D3V electrostatic parallel Particle-in-Cell (PIC) code BIT1 during ELM-free or no-ELM. As a first approximation plasma-surface interaction processes are not included in this model. The burning plasma conditions correspond to the ITER Q = 10, 15 MA baseline at q95 = 3, for which the poloidal length of the 1D SOL is ~20 m from inner to outer target. Typical upstream separatrix parameters of ne~3-5x10E19 m-3, Te~100-150 eV, and Ti~200-300 eV are assumed, guided by SOLPS-ITER code runs. Inclined magnetic fields at the targets of (~5) are included, as are the particle collisions, with a total of 3.4x10E5 poloidal grid cells giving shortening factors of 20. In these simulations, for the first time, the neutrals are included but the impurities are neglected. A typical simulation requires up to 60 days running massively parallel 1152-2304 cores of the EU Marconi super-computer.

13.09.2022 15:40 Poster session 1

Fusion – 1014

Towards refactoring of the TOKES tokamak plasma transient code

Leon Bogdanoviæ1, Sergey Pestchanyi2, Leon Kos3

1Faculty of Mechanical Engineering, University of Ljubljana, Aškerèeva 6, 1000 Ljubljana, Slovenia

2Karlsruhe Institute of Technology (KIT), Hermann von Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

3University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerèeva cesta 6, 1000 Ljubljana, Slovenia


The TOKES (“Tokamak Equilibrium and Surfaces”) code simulates numerically dynamics of the thermonuclear deuterium-tritium (D-T) plasma in ITER core, in scrape-off layer (SOL) calculates heat ?ux to the tokamak walls and heat transport inside the solid walls. It takes into account phase transitions of the wall material, i.e., tungsten (W), including vaporization. After vaporization start, TOKES simulates dynamics of vaporized W in vacuum vessel, its ionization and W-D-T plasma dynamics, including photonic radiation. The code features a numerical meshing out to all wall surfaces with the possibility of spatially variable grid resolution on the mesh. It includes standard surface interactions such as sputtering, but also surface vaporization and, importantly, a vapour shielding module. TOKES has been extensively used for several years in specific ITER studies, covering in particular simulations of disruption mitigation by massive gas and shattered pellet injection and the impact of heat fluxes due to non-mitigated disruptions and ELMs (Edge-Localized Modes), including vapour shielding effects. The code is also being applied to JET-ILW (ITER-like wall) and to the EU DEMO PFC design activities considering, in particular the impact of disruptions and sacrificial limiters. Compared with more conventional boundary codes, it has the advantage of rapid run times, permitting extensive parametric studies even at the reactor scale like required in the DEMO design phase. Developed over almost two decades at KIT, the source code of TOKES in Pascal is still compiled in Delphi, a commercial Integrated Development Environment (IDE), under Windows on single machines. For the benefit of TOKES preservation and availability to the fusion community refactoring of its source code to an open-source solution is needed. This paper presents the progress on TOKES code refactoring under Free Pascal on Linux as well as some examples of simulation results visualization in a prototype Graphical User Interface (GUI).

13.09.2022 15:40 Poster session 1

Fusion – 1015

Natural convection cooling in DEMO Vacuum Vessel during ex-VV LOCA

Martin Draksler1, Primož Črne1, Christian Bachmann2, Sergio Ciattaglia2, Ivo Moscato2, Boštjan Končar1

1Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Eurofusion – Programme Management Unit, Boltzmannstrasse 2, 85748 Garching, Germany


This study evaluates the temperature evolution of DEMO in-vessel components (IVCs) when cooling is impaired during an ex-vessel LOCA, including the 120s-long plasma ramp-down. It is conservatively assumed that a large (guillotine) break in the main feeding line of the BB PHTS causes an instant loss of cooling in all BB segments.
The boundary conditions and decay heat data have been set according to the latest update of DEMO design and conditions. Design upgrade includes BB supports that are attached to the cold VV wall. Concerning the heat transfer model, the radiation heat transfer is modelled explicitly using the S2S radiation model. Transient CFD analyses with the ANSYS Fluent code have been carried out to assess the temperature distribution of the in-vessel components and the redistribution of thermal loads due to in-vessel natural convection of injected helium, heat conduction through supports and radiation heat transfer.
The ex-vessel LOCA analysis described here shows that natural convection with filled gas provides a sufficient cooling mechanism for decay heat removal from BB segments even in case of an instant loss of cooling in all BB segments. The peak BB temperature occurs in the first hour from the onset of the event, when the maximum temperature in BB segments is around 550°C. The distribution of thermal loads in the BB segments shows that the natural convection cooling of the filled gas is an important contributor to decay heat removal.

13.09.2022 15:40 Poster session 1

Fusion – 1017

Beyond JET: the role of privately funded fusion research and first results from the ST40 high field spherical tokamak

Henri Weisen

Tokamak Energy Ltd, 173 Brook Drive, OX14 4SD, United Kingdom


The future path of publicly funded fusion development is well known and includes the international
ITER device and the demonstration power plant DEMO. The aim of this paper is present the role of
privately funded fusion research, as exemplified by Tokamak Energy Ltd. The past few years have
seen growing private sector support for developing commercial fusion reactors. These emerging
fusion startups, often in collaboration with the fusion research institutions from which they were
spawned, aim to fast track the development of commercial fusion with the objective of operating
concept-validation devices currently under design within this decade.
In a sign of things to come, ST40, the first high field spherical tokamak, operated by Tokamak Energy
Ltd, has recently passed an important milestone, achieving a ‘JET-like’ plasma ion temperature of 10
keV in a plasma volume ~100 times smaller than that of JET and a heating power 15 times below that
of JET. It is also the first magnetic confinement device with a plasma volume below 1m3 that has
achieved a fusion triple product nioTio?E =1019 m-3keVs (ion density × ion temperature × confinement
time) at fusion relevant ion temperatures. The triple product, introduced by fusion pioneer J.D.
Lawson, is an important metric for evaluating progress towards a reactor.
Tokamak Energy Ltd is undertaking the construction of a high field spherical tokamak making use of
high temperature superconductors (HTS) operating at a plasma current similar to that of JET (up to
4MA), albeit with a much smaller plasma volume, for the purpose of technology validation. This
device is to pave the way for an electricity generating prototype D-T reactor in the next decade

13.09.2022 15:40 Poster session 1

Fusion – 1018

Dose field due to activated cooling water in simple tokamak model

Igor Lengar, Domen Kotnik

Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Water is the cooling fluid of present large tokamaks and the candidate medium for future tokamaks like ITER and DEMO. During fusion reactor operation the water gets activated in the first wall due to reactions with neutrons. The resulting activated nuclei N-16 and O-19 emit gamma rays and the N-17 nuclei emit neutrons, half-lives of the reactions range between 4 and 27 seconds. The neutron flux and correspondingly water activation are largest when the water flows through the first wall. However, the majority of decay takes place at larger distances from the first wall, consequently a considerable amount of the decay radiation takes place also outside the biological shield.
The majority of studies is devoted to calculations of the dose at different positions during reactor operation. In the present study the dose, originating from the transported water, is evaluated. The study is performed with a combination of MCNP neutron transport for determination of water activation by plasma neutrons and calculation of the irradiated water transport in combination with its decay for determination of the secondary source term. The study is performed on a simplified model of a large tokamak in order to evaluate the simplified procedure as opposed to lengthy more detailed calculations.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1106

Analysis of Influence of DEC Equipment on Severe Accident Development

Matjaž Leskovar, Mitja Uršiè, Janez Kokalj

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


A hypothetical severe accident in the Krško NPP was analysed with the MELCOR code version 2.2 taking into account mitigation measures for heat removal from the containment solely by the design extension conditions alternative safety systems. As the initiating event, a strong earthquake was considered, resulting in a simultaneous station black-out and large break loss-of-coolant accident. The following scenarios were analysed: (1) no mitigation, (2) after reactor pressure vessel failure water injection through containment sprays, (3) after reactor pressure vessel failure water injection simultaneously through containment sprays and into the reactor coolant system, (4) after reactor pressure vessel failure water injection into the reactor coolant system, (5) after reactor pressure vessel failure water injection into the reactor coolant system, without operable alternative residual heat removal system heat exchanger but with operable alternative auxiliary feedwater system.

In the paper the five analysed severe accident scenarios and the applied Krško NPP MELCOR code model will be presented. The main focus will be given to the molten core concrete interaction. It turned out that the molten core concrete interaction in the reactor cavity can be stopped if the molten core is flooded soon after it is released from the failed reactor vessel. The extend of the molten core concrete interaction is very sensitive on the flooding time. The simulation results revealed that the heat transfer through the steam generators by natural circulation of the atmosphere in the failed reactor coolant system is not sufficient to stabilize the severe accident. In this scenario (5) the containment atmosphere was periodically released through the passive containment filtered venting system, like in the unmitigated case (1), whereas in the other mitigated scenarios (2-4) no releases occurred.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1107

Effect of sump pH on iodine release from VVER-1000/V-320 containment during LB LOCA

Adam Kecek, Lubomir Denk

ÚJV Øež, a.s., Hlavní 130, Øež, 250 68 Husinec, Czech Republic


Estimation of fission product source term during various loss-of-coolant accidents at nuclear power plants is the final step in the chain of NPPs safety analyses. The number of studied fission products is large, but one of them, thanks to its bioactivity and complex physical and chemical behaviour, stands out. Iodine behaviour within NPPs containments during various conditions has been studied worldwide in numerous research projects during past decades. One of them is the RTF series introduced within the OECD NEA BIP project, which aims at effect of sump pH on iodine revolatilization. The scope of this paper is to validate COCOSYS 2.4v5 code on this experiment and then transfer the knowledge to the analysis of LB LOCA at VVER-1000/V-320, where the effect of variable sump pH on iodine release will be studied. The study will aim at two different initial chemical speciation according to the R.G. 1.183 and R.G. 1.195 respectively.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1108


Julia Herrero-Otero1, Enrique Meléndez Asensio2, Miguel Sánchez Perea2, César Queral3, Marcos Cabezas1, Sergio Courtin1, Alberto García-Herranz1, Carlos París1, Rafael Iglesias1

1Universidad Politécnica de Madrid, Alenza 4, 28003, Madrid, Spain

2Nuclear Safety Council, C/Justo Dorado 11, 28040 MADRID, Spain

3Universidad Politécnica de Madrid (UPM), Nuclear Engineering Department, José Gutiérrez Abascal 2, 28006 Madrid, Spain


Oversight of the licensee’s performance should be done from an independent position. This position is best achieved when the regulatory body develops its own methodologies and tools. In particular, in the case of probabilistic safety analysis (PSA), large number of hypothesis and assumptions behind the model make it very difficult to perform sound regulatory analyses using licensee models. Therefore, the development of a PSA model to be used by the regulator provides a better understanding of the risks of nuclear power plants and an improvement of the regulatory practice, at the same time enhancing consistency of the risk perspective across regulatory actions.

In this context, the Spanish Regulatory Body (CSN), in collaboration with the Universidad Politécnica de Madrid (UPM), has been developing its own models: a standardized generic Standardized Plant Analysis Risk model (SPAR-CSN) for a PWR-WEC 3-loop design and a specific model of an individual Spanish NPP as well as the corresponding model of a PWR-WEC 3-loop plant. The development of both SPAR models has allowed analyzing, from a risk point of view, two actual incidents including sensitivity analyses. The results obtained have led to the identification of safety-critical systems and components.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1109

UQ Study Of The Effect Of Oxidation Parameters On SFP Accident Progression Using MELCOR 2.2.

Mateusz Malicki

Paul Scherrer Institut, Villigen, CH-5232 Villigen PSI, Switzerland


The Fukushima accident brought more attention to the Spent Fuel Pools (SFP) safety, including potential Severe Accident (SA) scenarios. Consequently, studies, code validations and experiments dedicated to SFP SA were intensified in past years to improve knowledge in that area. One of the main subjects is Fuel Assembly degradation as a result of cladding oxidation processes. In the case of SA codes, the implemented models are simplified, leading to increased calculation uncertainty and often forcing the conservative approaches. To decrease the uncertainty and improve the quality of the results, broad uncertainty quantification (UQ) study needs to be done.
The study presented here is done as a part of the WP6 of European Project MUSA in which the main goal is to develop a methodology for severe accident UQ. Establishing new methods required a deep understanding of the issue of cladding oxidation and a wide spectrum of analyses, including review of the phenomenology and different tools. In this task, the authors decided to concentrate on the uncertainties related to the cladding oxidation processes choosing fifteen uncertain parameters linked with oxidation and core degradation phenomena. As Figures Of Merit (FOM), onset time of Fission Product (FP) release and total FP released mass were chosen. The calculations were carried out using a severe accident code MELCOR2.2 within the SNAP environment and the input deck of the Fukushima unit 4 SFP. The uncertainty quantification tool DAKOTA supported by in-house PYTHON scripts was used for statistical analyses.
Preliminary analysis showed that in the early phase of the SA transient, the oxidation cutoff temperatures and radiative heat exchange factors are correlated with the FOMs. Based on the performed calculations, the authors underline the need for further investigation of the oxidation models parameters and FOM definition. The results of sensitivity calculations showed a significant effect of the input deck nodalization on the final UQ results, and thereby it is recommended to address this issue in future investigations.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1110

Supervised Machine Learning-based Feature Selection in the Frame of BEPU in Severe Accidents

Michela Angelucci1, Rafael Bocanegra2, Sandro Paci1, Luis E. Herranz2

1University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy

2As. CIEMAT, Av. Complutense, 40, 28040 Madrid, Spain


In the framework of Best Estimate Plus Uncertainty (BEPU) application to Severe Accidents (SA) – related scenarios, the assessment of the uncertainties linked to the simulation results is a mandatory step. However, Uncertainty Quantification (UQ) analyses do not provide an insight into the contribution of individual input parameters – hereon called “features” – to the calculated uncertainty band. To this end, a complementary sensitivity analysis is often needed. The reason for this is twofold: to identify the features with the highest influence on the addressed output response/s, and, in turn, to focus research on dominant features to effectively reduce uncertainty.
In this regard, the objective of the present work is to establish a data analysis methodology allowing a deeper understanding of the features driving the uncertainty. To do so, an alternative approach for sensitivity analysis, based on the application of supervised Machine Learning (ML), is proposed. The methodology intends to exploits various regression techniques, while facing two major constraints: the high number of features involved, the reason of which can be found in the intrinsic complexity of SA phenomenology, and the small size of the database, due to the computational cost of SA codes.
As a case study, the proposed ML-based approach is applied to a database developed in the MUSA H2020 project: data coming from the simulation of a simplified but still representative SA scenario, such as PHEBUS FPT1 test, with the MELCOR code are fed to different ML regression algorithms. Preliminary results show that the use of an appropriate algorithm can actually help in shrinking the number of features, thus improving the interpretability of the model and the identification of the variables that are responsible for most of the uncertainty on the response/s. Moreover, the study suggests the need of corroborating the results with the physical meaning. In other words, expert judgement might play a key role in both feature selection and in the understanding of the results from the sensitivity analysis.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1111

Application of dynamic bayesian network in initiating event calculation

Aleksej Kaszko1, Piotr Kopka2

1National Centre for Nuclear Research, ul. Andrzeja So³tana 7, Otwock-Œwierk, Poland

2Narodowe Centrum Badañ J¹drowych, Soltana 7, 05-400 Otwock, Poland


The Probabilistic Safety Assessment (PSA) technique uses Initiating Events (IE) probability as the initiator of accident progression in Event Trees (ET). There are various ways to calculate initiating events described in IAEA TECDOC-719. Each IE calculation method provides static probability, neglects aging, and doesn’t provide information regarding individual influencers on IE probability. Authors developed the technique through a Dynamic Bayesian Network (DBN) that evaluates the probability of initiating event from multiple hazards with aging consideration.
This paper presents a methodology and a structure of a Dynamic Bayesian Network for Loss of Offsite Power (LOOP) IE with such hazards as earthquakes and tsunamis. Earthquake-induced tsunami happed in 2011 in the cost of Japan due to one of the most powerful earthquakes called the Tohoku earthquake with 9.0 Mw which caused LOOP, and therefore this case could be representative for such structure creation.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1112

RELAP5 simulations of total loss of feedwater in a PWR

Andrej Prošek

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


In the Europe the design extension conditions (DEC) were introduced after Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The purpose of the study is to determine available time before core degradation and needed DEC safety features total loss of all feed water in a two-loop pressurized water reactor. In its documents, both WENRA (Western European Association of Nuclear Regulators) and the International Atomic Energy Agency (WENRA) present total loss of all feed water initiating event as a possible initiating event for existing power plants.

For simulations the U.S. Nuclear Regulatory Commission advance TRACE computer is used. The latest RACE Patch 2006 has been used. The TRACE input deck has been developed based on the RELAP5 standard input deck, which is verified and validated. The initiating event loss of all feedwater is multiple failure in which besides the loss of main feed water also auxiliary feedwater is lost. This system consists of two motor-driven and a turbine driven pump pump. It is also assumed the operator’s action to trip the reactor coolant pump in accordance with Emergency Operating Procedures (EOP). Initially the reactor is assumed to operate at 100% power. It is assumed that both High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection (LPSI) trains and batteries are available. The results section will show comparison of TRACE calculated results to RELAP5/MOD3.3 calculated results. First, TRACE steady state results will be shown, followed by the results of total loss of feedwater for determination of time available before core degradation. Finally, the simulated results of total loss of feedwater with design extension conditions safety features available will be shown.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1113

Simulations of FLOAT debris quenching experiments with the MC3D code

Tim Kelhar1, Markus Petroff2, Janez Kokalj3, Mitja Uršiè3, Rudi Kulenovic2, Matjaž Leskovar3, Leon Cizelj3, Joerg Starflinger2

1Fakulteta za matematiko in fiziko, Jadranska 19, 1111 LJUBLJANA, Slovenia

2Institute of Nuclear Technology and Energy Systems (IKE) University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart, Germany

3Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Severe accidents in nuclear power plants can result in the melting of a reactor core and the potential development of a debris bed. For relocated debris beds at the reactor cavity non-condensable gases can be generated by a possible MCCI (molten core concrete interaction). A safety-relevant question is whether the debris bed geometry is coolable.

Continuous computational and experimental research on debris bed cooling is being performed to investigate the cooling capabilities of different configurations. Several test facilities have been built. Recently, a series of experiments on the FLOAT test facility at IKE, University of Stuttgart, Germany, were performed. The debris bed coolability in cases of top-flooding with counter-current gas flow (air), injected at the bottom of the bed, have been investigated. The air injection gives the FLOAT test facility an option for simulating a non-condensable gas flow.

As part of our research, two FLOAT experiments, F.32 and F.33, with high initial debris bed temperatures (700 °C) will be simulated, of which one has an air injection. The simulation will be performed using the MC3D code (IRSN, France). The objectives of the paper are to consider 3-D effects in the simulations and to assess the impact of the non-condensable gases, for the top-flooding configuration, on the quenching of superheated particle beds.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1114

Multi-fluid simulation of PECA pool scrubbing experiments

Matic Kunšek1, Luis E. Herranz2, Ivo Kljenak1, Leon Cizelj1

1Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2As. CIEMAT, Av. Complutense, 40, 28040 Madrid, Spain


During a hypothetical severe accident in a light water reactor nuclear power plant, the fuel could melt and there is a possibility, that some of the radioactive material could be released as particles to the surrounding area. The releases of the radioactive material can be reduced with the application of pool scrubbing, where contaminated gases are filtered through a pool of liquid water. To understand what is happening during pool scrubbing, phenomena at the local scale need to be understood. Specifically, since the gases enter the scrubbing pool as a jet that disperses into bubbles, the behavior of the particle removal from the bubbles is crucial for understanding the pool scrubbing phenomena.
In the proposed paper, pool scrubbing PECA experiments, performed at CIEMAT (Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas) in Madrid (Spain) were simulated using multi-fluid modelling. Simulations were performed using the open-source Computational Fluid Dynamics code OpenFoam. The proposed approach considers only the decontamination in the rise region of pool scrubbing. In the modelling, four different fluids are considered: gas, liquid and two particle fluids (particle fluid 1 within gas bubbles and particle fluid 2 within liquid). All fluids are described in the Eulerian frame. The particles transport from bubbles to liquid during the bubble rise in the scrubbing pool is simulated as a transfer via a subgrid model from particle fluid 1 to particle fluid 2, based on the simulation of particle motion within bubbles. Namely, the subgrid model takes into account that, due to bubbles rising, the inner air motion moves particles inside bubbles (particle fluid 1) due to interfacial drag. The particles first migrate towards the bubble surface and then out into the liquid (to become part of particle fluid 2). The particle densities and bubble diameters were prescribed, based on data from the literature. The simulation results were analyzed and the decontamination factor, which is the resulting measure of the scrubbing efficiency, was calculated and compared with experimental measurements.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1115

Simulation of fuel-coolant interaction in stratified configuration in reactor geometry

Janez Kokalj, Mitja Uršiè, Matjaž Leskovar

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


During a hypothetical severe accident in a light water nuclear power plant, the molten reactor core may come in contact with the coolant water. One of the consequences can be a vapour explosion. Typically in nuclear safety, the vapour explosions are mostly analysed in the melt jet-coolant pool geometry. In the stratified melt-coolant configuration, which was less analysed, a layer of melt spreads below a layer of coolant. The stratified melt-coolant configuration was believed to be incapable of producing a strong energetic interaction between the melt and coolant. This belief was based on the conclusions from the past theoretical and some experimental research with simulant materials, where interfacial instabilities of the melt were not observed and consequently the interface was almost flat, yielding a strong hypothesis of the subsequent models.
However, the results from recent experiments performed at the PULiMS and SES facilities (KTH, Sweden) with corium simulants materials contradict this hypothesis. In some of the tests, a premixed layer of ejected melt drops in water was clearly visible and was followed by strong spontaneous vapour explosions. Based on the past experimental and analytical research, model for the melt-coolant premixed layer formation in stratified configuration was developed.
The purpose of our research is to improve the knowledge, understanding and modelling of the fuel-coolant interaction and vapour explosion in stratified configuration in reactor-like conditions. Therefore, simulations with MC3D code (IRSN, France) will demonstrate the model’s capability to describe the premixed layer formation and followed vapour explosion in reactor geometry. The analyses will be performed to assess the possible pressure loads on the reactor cavity during vapour explosions, which is of high interest in nuclear safety.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1116

Risk-Informed Decision Making: A Showcase Approach or An Alternative Solution For Safety Assessment in NPPs

Seyed Ali Hosseini1, Reza Akbari2, Amir Saeed Shirani1, Francesco Saverio D’Auria2, Alireza Najafi1

1Faculty of Engineering, Shahid Beheshti University, Daneshjoo Blvd, Evin, 1983963113 Tehran, Iran

2University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy


During the last decades, technology has helped human beings to simulate environmental phenomena and achieve meaningful information. Nevertheless, this achievement necessarily does not mean that we are entirely able to predict anything around us. The risk assessment as a quantitative-qualitative tool may provide us a picture of dedicated accidents, but this sight is a blurred picture in case of unknown accidents. Today, risk assessments in NPPs are mainly performed based on two independent approaches, Probabilistic Safety Assessment (PSA) and Deterministic Safety Assessment (DSA). PSA is a proper tool for identifying event sequences leading to a damaged state and evaluating the corresponding probability. Conventionally, PSA is considered a complement to the traditional DSA. While these approaches are completed each other’s with interconnections, there is not any integrated insight about how to make a consistent decision. In other words, the interpretation of results from DSA and PSA builds parallel decision-making about considered scenarios. The IAEA INSAG-25 document proposed an integrated process for Risk-Informed Decision-Making (RIDM) in NPPs. The core of the RIDM uses several options (like PSA insights, deterministic consideration, etc.) to provide a balanced input to decision-makers on how the principles are met. All previous studies about RIDM mentioned the weighting of options against principles. However, there is no precise formulation for decision-making. In other words, decision-making based on options is a deliberative process between decision-makers. Thus, the RIDM framework (as a showcase approach) can still have implementation challenges that need to be addressed in the future. Also, understanding the impact of uncertainties and quantifying them is another challenging issue for the RIDM framework. The current work reviews the status of RIDM practical implementation and issues about further development.

13.09.2022 15:40 Poster session 1

Severe accidents and PSA – 1117

Calculation of the QUENCH-02 experiment with the ASYST code including the uncertainty evaluation

Siniša Šadek, Renato Pavlinac, Karlo Ivanjko, Davor Grgiæ

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia


Uncertainties in the safety assessment of nuclear facilities are the result of random variability in the values of the variables describing the system and its initial and boundary conditions, and the lack of precision of calculation models. Varying the input variables will cause the results of the calculation to be scattered depending on the predefined confidence level. The ASYST code, developed as part of an international nuclear technology ASYST Development and Training Program (ADTP) managed by Innovative Systems Software (ISS), is used to perform an uncertainty analysis of the QUENCH-02 experiment conducted at the Karlsruhe Institute of Technology. The code uses a probabilistic methodology based on the propagation of input uncertainties.
The QUENCH facility consists of 21 fuel rod simulators, 20 of which are electrically heated. The aim of the QUENCH experiments is to examine hydrogen source term and the behaviour of the fuel rod cladding during core reflood, in a typical PWR reactor. The QUENCH-02 experiment did not have a pre-oxidation phase, thus the objective was the investigation of PWR fuel rods behaviour with little oxidation. The experiment consisted of a heatup phase to temperature of 900 K, a transient phase in which the temperature rose to approximately 2300 K, and a quenching phase with mass flow of water 40-50 g/s.
For selected input parameters, such as steam/water flow, electric power and other relevant boundary conditions, it is necessary to define their probability density functions. Input databases are then prepared for individual calculations based on the selected level and confidence interval. The number of calculations is large enough to ensure at least 95% coverage of expected output results and uncertainty limits. The results of the calculations are compared with the experimental measurements. The Pearson correlation coefficient is used to obtain correlation between the input uncertain parameters and the output data (mass of hydrogen, temperature of the heated rods and the level of their damage).

13.09.2022 15:40 Poster session 1

Emerging reactor technologies – 1201

Nuclear power for the decarbonization of the district heating sector in the Czech Republic

David Mašata1, Hussein Abdulkareem Saleh Abushamah1, Jana Jiøièková2, Radek Škoda2

1University of West Bohemia, Faculty of Electrical Engineering, Univerzitni 8, 301 00 Pilsen, Czech Republic

2University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic


The article summarizes and evaluates the progress of research on nuclear power utilization for decarbonizing the district heating sector. Nuclear power represents the most concentrated carbon-free energy source, which should become an integral part of the heat energy industry of the future. The current primary fuel mix for district heating is based mainly on fossil fuels – coal and gas. This work evaluates the current situation of nuclear-based district heating by conventional large nuclear power plant units and maps the future possibilities of new small modular reactors concepts applications or new-build plants integration. The Czech nuclear heating concept TEPLATOR is introduced and compared with other SMR concepts for heating. Finally, a summary of all possibilities of nuclear source applications for the Czech Republic district heating sector decarbonization is provided.

13.09.2022 15:40 Poster session 1

Emerging reactor technologies – 1202

Special experimental environment for Gen. IV reactors with graphite reflector

Eva Vilimova, Tomáš Peltan, Radek Škoda

University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic


Nowadays, there is an increasing demand for new SMR reactors with a wide range of applications, often classified as a new generation IV. reactors. Unfortunately, there is no operating nuclear reactor meeting the characteristics of Gen. IV reactors by its technical design and features. Generation IV nuclear reactors are intensively developed worldwide, including the Czech Republic. At least two general Gen. IV thermal neutron reactor concepts use graphite as a moderator or reflector, as do many concepts from the very popular small modular reactors. To support research activities linked with the development of these reactors, an appropriate experimental environment and resources simulating conditions expected in Gen. IV reactors with graphite are needed. The calculated data confirm the results obtained during previous research. The experiment at LR-0 with a graphite reflector gives better results of neutron flux distribution in the reflector due to the extra graphite reflector layer and central graphite plugs. Besides, the core arrangement is included in a set of experiments supporting the research of reactor cores with graphite reflectors. The main reason for this article is to support the development of a new functional sample of neutron instrumentation for Gen. IV reactors.

13.09.2022 15:40 Poster session 1

Emerging reactor technologies – 1203

The Integration of Hybrid Nuclear Renewable Energy Systems

Jan Lokar, Robert Bergant, Klemen Debelak

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia


The two principal options for low carbon energy generation are nuclear energy and renewable energy. The amount of renewables in energy systems is increasing. As the decarbonization of energy generation is one of the main goals of the sustainable development, many activities are being carried out in the field of synergy operation of the low carbon energy generation sources.

Hybrid nuclear and renewable energy systems are defined as integrated systems consisted of nuclear reactors, renewable energy generation sources and industrial processes that can simultaneously achieve grid flexibility, greenhouse gas emission reductions and optimal use of investment capital. Those systems can take advantage of coupling nuclear and renewable energy generation sources to increase the benefits of each technology to provide reliable, sustainable electricity to the grid and to provide low carbon energy to other industrial energy sectors as well.

In the recent years there are many research activities in this field. Integration of these systems is currently in development phase. There are some loosely coupled hybrid nuclear and renewable energy systems in operation with primary focus of supporting electricity demand. As these systems presents the potential for sustainable energy generation, a higher development is expected in the future.

In this paper the basic concept of hybrid nuclear and renewable energy systems is presented. In the second part, an overview of current research activities is carried out. In the final part of this paper an example of hybrid system of GEN group is presented.

13.09.2022 15:40 Poster session 1

Emerging reactor technologies – 1204

Development of the MATLAB Tool for Optimization of the Nuclear Hybrid Energy System Dedicated for Hydrogen Production

Piotr Darnowski, Wojciech Kubiñski

Warsaw University of Technology, Pl. Politechniki 1, 00-661 Warsaw, Poland


The paper presents the development process, features and example application of the numerical tool for optimization of Hybrid Energy System (HES) with nuclear reactors and dedicated to hydrogen production. The tool was developed in the MATLAB environment, and it applies the Mixed Integer Genetic Algorithm to obtain a HES configuration fulfilling predefined technical, economical and other constraints.
The work was focused on large scale systems with nuclear reactors and coupled with various hydrogen production technologies. The internal database of possible technologies, their performance and type of series of possible units was created on the basis of available literature. The work considers HES systems that utilize different nuclear reactor technologies like Gen-III and Gen-IV reactors and different hydrogen production technologies, including Low Temperature Electrolysis (LTE) and High Temperature Electrolysis (HTSE) and thermo-chemical cycles.
The example application of the tool was presented for the HES with minimal production of hydrogen ~100 000 t/year. Several constraints were considered, including the cost, deployment time and technology readiness level.

13.09.2022 16:20 Materials in nuclear technology

Materials in nuclear technology – 601

Long-term operation safety performance of nuclear civil engineering structures – corrosion aspects

Nina Gartner, Miha Hren, Tadeja Kosec, Andraž Legat

Zavod za gradbeništvo Slovenije, Dimièeva 12, 1000 LJUBLJANA, Slovenia


Concrete structures in nuclear power plants (NPPs) provide a foundation, structural support, biological shielding, containment and protection. If not controlled, the ageing of these concrete structures could increase the risk to public health and safety. The EU-funded (H2020) project ACES ( investigates the ways to advance the current assessment of safety performance for the long-term operation of nuclear power plants. Different experimental and modelling techniques are used to study deterioration and ageing mechanisms, such as radiation effects, internal swelling reactions and liner corrosion. The findings will assist in improving the safety and operational designs of next-generation nuclear plants.
Concrete structures with embedded steel liners are a considerable part of the NPP’s infrastructure. Although concrete enables passivation of the steel surface due to its alkalinity, many corrosion processes occur due to environmental conditions (e.g. chlorides) and geometry on the steel surface (e.g. crevice). Many of these corrosion processes are hidden inside the concrete until corrosion propagates to the stage where mitigation can already be very complex and costly. Two of the main identified corrosion concerns in NPPs are 1) chloride-induced corrosion of steel cylinder concrete pipes (SCCPs) and 2) crevice corrosion of steel liner embedded in the concrete of containment building.
In the case of SCCPs, the corrosion mechanisms are well known and the ACES project is focusing on suitable non-destructive inspection techniques (NDTs), which could be implemented on the mobile robotic platform and enable early detection of corrosion inside the pipes. In this study, two techniques were evaluated, i.e. half-cell potential mapping and galvanostatic pulse technique. Such equipment is commercially available, cost-effective, portable, robust, and relatively easy to use. Laboratory investigation on mock-up specimens showed that the half-cell potential mapping is a valuable method for fast qualitative corrosion risk estimation, showing weak spots on the steel liner surface. The galvanostatic pulse technique is slower, but in addition to half-cell potential, it can also provide a quantitative indication of corrosion rate.
Unlike for SCCPs, almost no information is available in the literature on the crevice corrosion phenomenon of steel liner embedded in the concrete of containment buildings. Therefore, corrosion mechanisms that take place in containment buildings were studied through various laboratory tests. The most typical environmental and geometrical characteristics were used for laboratory set-ups to mimic interface deviations between steel liner and the concrete (i.e. air or wood crevice formers) from the NPP environment. Two corrosion monitoring techniques were used to study this phenomenon: Electrical resistance (ER) sensors and coupled multi-electrode array (CMEA) sensors. The advantage of these techniques is their ability to detect corrosion initiation, follow spatio/temporal distribution of corrosion currents and distinguish between different types of corrosion. X-ray computed microtomography (?XCT) was used to validate the measured results. The results indicate some unexpected correlations between crevice geometry and corrosion processes, which represent important input data for long-term phenomenological corrosion modelling of such systems, and consequently optimization of the existing methodology for ageing management of NPPs.

13.09.2022 16:40 Materials in nuclear technology

Materials in nuclear technology – 602

PATHFINDER: A tool for calculating pathways for fusion-activated materials

Priti Kanth, Mark Gilbert

United Kingdom Atomic Energy Authority, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom


The DT fusion reaction will produce 14.1 MeV neutrons. These high-energy neutrons will interact with the surrounding material producing radioactive isotopes. Inventory codes are used to calculate the change in material composition and radiation fields produced from these irradiated materials. An isotope in the neutron environment can transmute to produce other isotopes, these next-generation isotopes can undergo nuclear reaction or radioactive decay producing a complex tree-like structure with branches, cross-linking, and loops. One of the most important results of such codes is the pathways to produce radioisotopes. Some inventory codes may lack the ability to calculate these pathways or in many cases, they may not have certain required features. This tool probes the nuclear data and reads the reaction cross-section and decay data into a coefficient matrix. And for a given initial isotope this tool can then create an activation tree using the linear chain method. It can also be used to calculate the reverse tree, all the parents from which the isotope can be created. Such a tool can be helpful in conjunction with the inventory codes to identify the source of the radioactive isotope or the different isotopes that can be created from a nuclide when kept in a neutron environment.

13.09.2022 17:00 Radioactive waste

Radioactive waste – 701

Deep Borehole Repository of HLW and SF – State of knowledge by SITEX.Network

Nadja Železnik1, Muriel Rocher2

1Elektroinštitut Milan Vidmar, Hajdrihova 2, p.p. 285, 1001 Ljubljana, Slovenia

2Institut de radioprotection et de surete nucleaire (IRSN), 60-68 Avenue de General Leclerc – BP17, 92262 FONTENAY-AUX-ROSES CEDEX, France


Many countries develop geological disposal projects for high-level radioactive waste (HLW) and/or spent fuel (SF) when considered as waste. The most widely selected option is the deep geological repository (DGR) concept, a mined repository with galleries located underground in geological layers into which packaged waste would be placed; the sites for such DGR have been selected in Finland, France and Sweden, and a site selection process is on-going in several other countries, such as in the United Kingdom, Germany and Switzerland.
As an alternative concept to the DGR, the deep borehole repository (DBR) concept, where waste packages are placed into single boreholes, relies on a similar safety strategy: confining and isolating the waste from the biosphere and surface natural phenomena in order to respectively rely on the geological environment to ensure long term passive safety and reduce the risk of human intrusion. The concept of DBR was first considered in the 1950s, but was rejected until the 2000s as it was far beyond existing drilling capabilities among others, given the constraints for HLW and SF management.
New technical developments in the drilling field relaunched the interest of a safe management of HLW and SF based on DBR concept in several countries. Therefore, the SITEX.Network association developed an overview of the existing studies that have been published on the DBR concept with information on the concept itself, on deployment strategies and methods, on issues associated with requirements related to waste packages and borehole equipment, hydro-geology, disposal operation, backfilling and sealing, and finally on safety analyses. The main aim is to provide bibliographical overview providing the state of knowledge about the DBR concept, the technical solutions for its implementation or major obstacles evidenced as a basis to identify safety issues important to deal with in a Safety Case.
This could be considered to identify for the future R&D as well. This paper discusses also the controversial issue of DBR trying to provide information from different viewpoints, like the design options, R&D programs required, societal concerns and regulatory needs.

13.09.2022 17:20 Radioactive waste

Radioactive waste – 702

Latest Development on Mobile Equipment for Nuclear Waste Treatment and Decommissioning

Hauke Grages, Celine Boulet

Framatome GmbH, Paul-Gossen-Str. 100, 91052 Erlangen, Germany


During operation of a nuclear power plant (NPP), parts and components have to be maintained, repaired or exchanged. In most cases the components and parts have to be decontaminated before repair or maintenance to reduce the dose rate. Liquid waste generated during operation or decontamination needs to be conditioned for final storage. During decommissioning, parts have to be decontaminated as well in order to reduce the dose rate to personal and enable final storage in a lower and thus cheaper activity category. The decontamination, conditioning as well as decommissioning equipment is only working periodically and is in standby mode most of the time. The decommissioning equipment needs flexibility to adapt to the particular existing on site conditions and its specific equipment to be decommissioned. The waste generated during the different decontamination and decommissioning activities needs to be conditioned on site according to its particular characteristics in order to be transported and put in final storage. Therefore, in terms of space, cost, installation, flexibility and transport, mobile equipment is favored by the NPP over stationary equipment for the decontamination, conditioning and dismantling fields, for both solid waste and waste water treatment. In particular, mobile equipment allow:
• The use of a single equipment for different buildings (for example for decontamination, cleaning, maintenance, or decommissioning);
• To be stored in another area with more available space;
• To be rented for the necessary time period by the NPP;
• A quick and easy set-up since the material is pre-assembled and pre-mounted;
• A modular design that can be configured so that parts can be removed, replaced or added easily;
• An easy transport in a container with no expensive special transport;
• In the case of decommissioning for example, to be employed in an environment with limited auxiliary equipment available, since everything needed is in the container.
The module contains the necessary connections (electricity, water, chemicals etc) and is ready for operation as soon as connected to the plant. Depending on the size, the mobile equipment can be composed of a single or several modules transported separately in containers and connected on site. The equipment is pre-assembled to ensure a quick and easy set-up.
This presentation provides an overview of different state of the art mobile equipment available on the nuclear market as well as within Framatome in the fields of decontamination, waste conditioning and decommissioning.

13.09.2022 17:40 Radioactive waste

Radioactive waste – 703

Development of an evaluation method for mobile radioactive contaminants for assessing public exposure risk in accidental events during decommissioning of nuclear power station

Tsuyoshi Sasagawa, Taro Shimada, Seiji Takeda

Atomic Energy Agency of the R.K., Republic Square 13, 480013 ALMATY, Kazakhstan


Nuclear regulatory inspection system was introduced in Japan, inspections with risk-informed has been conducted for operation of nuclear facilities. While it has been confirmed for nuclear power stations undergoing decommissioning that the public exposure dose in an accident with maximum release of radioactivity is less than 5 mSv in the approved decommissioning plan, no methodology has been established to identify the inspection targets in regulatory inspections. The significance of the inspection targets is not determined only by the initial radioactive inventory, but the risk is expected to change with the progress of decommissioning due to the non-routine dismantling operations. Therefore, it is necessary to develop a method to quantitatively assess the temporal and spatial changes of radiation risk and evaluate the relative significance as a target of regulatory inspection during decommissioning of nuclear power stations. Radioactive particles are generated and dispersed into air by cutting of metal and concrete components. The particles collected at such as the filters of local and building may be released at accidents such as fire and explosion. These contaminants are called by mobile radioactive contaminants, hereinafter called the mobile contaminants. As a first step in the development of the risk assessment method, quantitative assessment method with dismantling progress of the mobile contaminants changing temporally and spatially and governing the public exposure dose during an accident was developed in this study.
The initial radioactive inventory of the nuclear power plant required for the assessment of the mobile contaminants is based on the equipment of reference BWR described in the NUREG/CR-0672, and the decommissioning process is based on the Japanese standard dismantling schedule. When the initiating event is assumed to be a fire, the mobile contaminants that could be released in the fire event are assumed to be radioactive particles accumulated in the filters of local and building during dismantling activity in the normal situation before the accident. The amounts of accumulated radioactivity in these filters were evaluated for each dismantled equipment based on the cut line, the migration rate into the air of the cutting methods (underwater or in air) and the like.
As a result, the amount of the mobile contaminants in-air cutting was shown to be about two orders of magnitude larger than that in underwater cutting for the same component. Some dismantling components and processes in which the mobile contaminants are larger for in-air cutting of equipment with low radioactivity were obtained, because underwater cutting is used only for components with high radioactivity such as reactor internals. This result indicates that it is necessary to evaluate not the initial radioactive inventory but the mobile contaminants generated during dismantling in order to select inspection targets that take into account the mobile contaminants that may be released during an accident. In addition, the replacement and transportation of the filters that reached the upper limit of filter accumulation caused spatial and temporal changes in the mobility inventory. Thus, it is important to understand the spatial and temporal profile of the mobile contaminants based on the working conditions of decommissioning process in order to consider the target of regulatory inspection.

14.09.2022 08:30 Invited lecture 2

Nuclear energy and society – 800

The challenges of future nuclear power in Europe & France

Abderrahim Al Mazouzi

Electricite de France, Research and Development Division, Avenue les Renardieres, Ecuelles, 77818 Moret sur Loing Cedex, France


The sustainable nuclear energy technology platform (SNETP) became an association (AISBL – international association) under Belgian law since 2019. It is based on 3 pillars – NUGENIA, ESNII and NC2I with the objective to drive innovation, develop research and innovation agenda, foster collaboration between its members and beyond, support knowledge transfer and European competitiveness.
SNETP is composed of more than 120 members from 25 countries, gathering nuclear power plant operators with research centers, nuclear industry, academia, SMEs and technical support organizations.
SNETP’s goal is to provide solutions to tackle the huge challenges facing all of us now and will be even more dramatic for the generations to come if we do not do anything about it right now. In fact, the Nuclear sector needs to highlight its resilience in this historic moment both from the sanitary view point (the covid is getting even stronger now a days) and also from the climate side.
EDF R&D, and active member of SNETP believes that nuclear technology is more than just a power supplier: it plays an indispensable role in the medical sector, particularly in terms of diagnosis and treatment of cancer (thus supporting Europe’s Beating Cancer Plan), as well as in industry, space, agriculture, etc. Nuclear is also a key place of innovation for new digital applications (such as artificial intelligence, block chain, IOT,..) as well as for operation and security of the European electrical grid together with variable renewables and storage.
This presentation will address briefly the role of SNETP within the European landscape and highlight some of the topics that are addressed by EDF R&D in collaborative manner.

14.09.2022 09:10 Nuclear energy and society

Nuclear energy and society – 802

Regulatory Requirements for Siting of New Slovenian NPP

Barbara Vokal Nemec, Benja Režonja Gumpot, Tomi Živko, Tomaž Nemec

Slovenian Nuclear Safety Administration, Litostrojska ulica 54, 1000 Ljubljana, Slovenia


Nuclear energy, as a low-carbon source, is an important contribution to reduction of the levels of CO2 emissions. The planning and preparation project to build a second nuclear reactor unit in Krško, the JEK 2, already started 15 years ago, at a site close to existing plant in Krško. The new project of JEK2 obtained the first permit called energy permit in July 2021. In January 2022 the documentation for initiative for the national spatial plan was prepared.

Licensing of new reactor facilities is under responsibility of the SNSA. The SNSA actively responded to the investor’s activities. A project team was formed at the SNSA with the aim to prepare the SNSA for such extensive task of licensing the new NPP. The regular exchange of information started between the investor GEN Energija and the SNSA to facilitate the preparation of licensing documentation and immediately address any open issues about the evaluation of the site and the design of the JEK2.

The SNSA activities are aimed at increasing its capabilities by employing new personnel and preparing the qualifications criteria as well. The Slovenian legislative framework was also upgraded with new revisions of acts on spatial planning, environmental protection and construction. The SNSA prepares amended regulations with nuclear safety requirements based on WENRA Safety Reference Levels and the IAEA requirements.

SNSA started with preparation of requirements for site evaluation according to the national regulations and the available IAEA standards. An initial safety analysis report shall be prepared that would include the basis for the site authorization. At the stage of the initial safety analysis report, information about the nuclear power plant design will be limited to a general design envelope, while information about the site is likely to be comprehensive. The article will describe the regulatory activities in preparation of requirements for site evaluation, necessary for further licensing.

14.09.2022 09:30 Nuclear energy and society

Nuclear energy and society – 803

Development of Indicators to Measure the Social Impacts of the Nuclear Power Plant Closure and Decommissioning

Hana Gerbelova, Ruth Shortall

Joint Research Centre/European Commission, Westerduinweg 3, NL-1755 LE Petten, Netherlands


This study presents results from a stakeholders meeting which seeks to identify possible indicators for evaluating and forecasting the social impacts resulting from the announcement of the closure of a nuclear power plant (NPP). Independently of the plans for nuclear energy at national level, each NPP will eventually close and enter the decommissioning phase. From a local point of view, the closure of such an important industrial facility may have major negative social impacts, especially in remote locations where NPPs are the main source of local employment and income.
It is vital for the long-term prosperity of the community that all relevant social impacts are identified in advance, so that the closure strategy focuses on effectively mitigating the negative impacts and increasing the potential benefits. However, the extent of the consequences in each hosting community will vary from one site to the other and this means that important social impacts are often overlooked in more generic assessments. Development of specialised indicators for the social impacts of the closure and decommission of NPPs could therefore provide effective support to measure the effects and progress of the social aspects related to these activities.
Following the principles of Social Impact Assessment (SIA), a participatory approach ensures that the social indicators are developed together with the relevant stakeholders and are based on their own experience. Such indicators allow monitoring changes over time in quantitative or qualitative terms. The indicators proposed in this study examine the impacts on the plant employees as well as the hosting community and cover various issues such as local disruption during decommissioning activities, temporary nature of decommissioning workforce, loss of business activity and future development of the site, provision community employment, human rights, human health and safety issues. The participatory approach of this study is complemented by a literature review to identify potentially appropriate indicators. The overall aim of this study is to contribute to a better understanding of consequences of closing the NPPs on the hosting community. The assessment can capture wider considerations that may contribute to improving the development of decommissioning plans in the future, including the dismantling of the nuclear facility, the disposal of any radioactive material and remediation of the soil, as well as other post-closure opportunities arising considering existing capacities and knowledge.

14.09.2022 14:50 Nuclear energy and society

Nuclear energy and society – 804

Role of Nuclear Energy in Achieving Decarbonisation Targets: Revisited

Ana Staniè

E&A Law, 42 Brook Street, London W1K 5DB, United Kingdom


In this year’s presentation I propose to revisit the role of nuclear energy in achieving decarbinisation targets in EU, UK and more broadly, a topic I discussed at NENE last year, given the recent legislative and other developments.
In particular, I propose to discuss (i) what the adoption of the Complimentary Climate Delegated Act by the European Commission in February 2022, the REPower EU Commission’s policy documents of March and May 2022 as well as the forthcoming vote in the European Parliament means for the role of nuclear energy in the EU; (ii) the recently adopted Energy Strategy in the UK means for the role of nuclear in the UK; and (iii) the fallout from the war in Ukraine and the sanctions adopted in response mean for nuclear energy more broadly.

14.09.2022 10:50 NPP operation

Nuclear power plant operation – 902

Reliability Analyses of Emergency Power Systems with focus on Multi-Group Diesel Configurations

Manuel Gueldner, Mariana Jockenhövel-Barttfeld

Framatome GmbH, Paul-Gossen-Str. 100, 91052 Erlangen, Germany


This paper focuses on the reliability analyses of Emergency Diesel Generator (EDG) systems in nuclear power plants and describes possible Multi-Group Diesel (MGD) configurations as a new concept to provide emergency power supply.
In Multi-Group Diesel configuration, several smaller qualified diesel units operate together to replace one large diesel unit. This principal is also used in the industrial sector, especially for higher power ranges.
Further to other technical issues which have to be considered in the general MGD concept and design, the reliability of the system is a very important aspect which has to be investigated.
Reliability analyses are performed as part of the safety demonstration process during the EDG design phase. It is used for overall modernization topics, as well as in new-build projects. The main objective of these analyses is to demonstrate that the reliability targets (the so-called safety goals) imposed on the EDG systems by the customer, regulators, or design authorities are fulfilled. Reliability analyses also aim to identify major contributors that lead to EDG unavailability. Based on this, design improvements are proposed, or even mandatory, to increase the reliability of the EDG design.
One of the main challenges of EDG reliability analyses during licensing is that they are primarily concerned with the reliability assessment of the digital instrumentation and control (I&C) systems, which are nowadays a common part of the EDG control and protection system.
The Framatome methodology uses fault tree modelling to estimate the failure probability of the EDG on demand (unavailability), given by the failure to start or to operate the EDG system.
The scope of the analysis includes failures of the EDG as well as their mechanical, electrical, and I&C support systems. The fault tree model includes independent failures and common-cause failures (CCF) of mechanical, electrical and digital I&C components leading to the failure to start or to operate the EDG.
Reliability analyses are considered design guidance for analyzing different MGD configurations from a reliability perspective, with the main objective of maximizing the reliability of the design. Sensitivity analyses are also conducted to investigate the impact of reliability parameter variations on the Multi-Group Diesel reliability.
The complete paper will therefore describe the reliability analyses of EDG systems in general and then focus on the Multi-Group Diesel configuration, highlighting the increased reliability of this structure and further advantages of the new MGD concept.

14.09.2022 11:10 NPP operation

Nuclear power plant operation – 903


Sergio Courtin1, César Queral2, Alberto García-Herranz1, Marcos Cabezas1, Julia Herrero-Otero1

1Universidad Politécnica de Madrid, Alenza 4, 28003, Madrid, Spain

2Universidad Politécnica de Madrid (UPM), Nuclear Engineering Department, José Gutiérrez Abascal 2, 28006 Madrid, Spain


One of the key lessons learned from the Fukushima Dai-ichi accident was the importance of the challenge presented by the loss of safety-related systems after a beyond design basis event (DBE). In particular, extended loss of power (ELAP) can severely compromise key safety functions associated with core cooling and containment, ultimately leading to reactor core damage.

FLEX is a set of diverse and flexible mitigation strategies that reduce the risks associated with beyond DBE conditions (NEI 12-06, 2012). These are based on the incorporation of various improvements, including portable equipment (diesel generators, diesel pumps, etc.) that provide multiple means of power supply and water supply to support key safety functions.

This paper quantifies the impact of FLEX strategies by incorporating them into the probabilistic safety assessment level 1 (L1PSA) of a generic standardized model (3-loop, PWR-WEC) developed by Universidad Politécnica de Madrid (UPM). As a result, a decrease in core damage frequency (CDF) and, therefore, an improvement in plant safety can be seen.

14.09.2022 11:30 NPP operation

Nuclear power plant operation – 904

Advancements in ultrasound inspection of nuclear power plants using robust transducers and deep learning

Željko Rapljenoviæ1, Marko Budimir1, Luka Posiloviæ2, Duje Medak2, Marko Subašiæ2, Fran Milkoviæ2, Sven Lonèariæ2

1INETEC-Institute for Nuclear Technology, Dolenica 28, 10250 Zagreb, Croatia

2University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia


The safety of nuclear power plants has always been one of the most important security issues in the industry in general. Numerous standards, techniques, and tools have been developed to deal specifically with the safety of nuclear power plants – one has specialized probes, robotized systems, electronics, and software. Although seen as a mature (or slowly evolving) industry, this notion about nuclear safety is a bit misleading – the area is developing in many promising new directions. Some recent global events will speed up this development even more.

On the other hand, the industry is currently going through digital transformation, which brings networking of devices, equipment, computers, and humans. This fourth industrial revolution promises speed, reliability, and efficiencies not possible up until now. In the NDE sector, new production techniques and traditional manufacturing lines are getting to be lights-out operations (near-total automation). The same is most probably going to happen with the safety inspections and quality insurance. Robotics and automation are improving worker safety and reducing human error. The well-being of inspectors working in hazardous environments is being taken care of. Most experts agree that the digitalization of NDE offers unprecedented opportunities to the world of inspection for infrastructure safety, inspector well-being, and even product design improvements. While the community tends to agree on the value proposition of digital transformation of NDE, it also recognizes the challenges associated with such a major shift in a well-established and regulated sector.

The work presented in this paper shows a part of the project that aims to develop a modular ultrasound diagnostic NDE system (consisting of exchangeable transducers, electronics, and acquisition/analysis software algorithms), for applications in hazardous environments within nuclear power plants. The paper will show how the software part of this system can reach near-total automation by implementing various deep learning algorithms as its features and, then, testing those algorithms on laboratory samples, showing encouraging results and promises of online monitoring applications.

Furthermore, future general prospects of this technology are discussed, and how this technology can affect the well-being of nuclear power plant inspectors and contribute to overall plant safety.

14.09.2022 11:50 NPP operation

Nuclear power plant operation – 905


Kaja Zupancic1, Jure Jazbinšek1, Tomaž Žagar2

1GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia

2Društvo jedrskih strokovnjakov Slovenije, Jamova 39, 1001 LJUBLJANA, Slovenia


Paper will present currently available processes for hydrogen and synthetic methane production at nuclear power plants. In addition, an overview of possible new technologies for production with future nuclear power plants will be given in the paper. A comparison of several fossil fuel free (carbon-free) hydrogen production technologies will be presented. Paper will give also cost analysis of various hydrogen production installations based on current market availability, price estimations, efficiencies and overall capacity factor.

14.09.2022 12:10 NPP operation

Nuclear power plant operation – 906

Unsupervised ML anomaly detection approach for NPPs LTO Program

Salvatore Angelo Cancemi, Rosa Lo Frano

University of Pisa, Department of Civil and Industrial Engineering, Largo Lucio Lazzarino 2, 56124 Pisa, Italy


Recent advancements in data-driven analysis methods, represented by those in artificial intelligence and machine learning, are improving the NPP performance ranging from the anomaly detection to the automated operational control of its complex systems. Indeed, the application of these methods can significantly improve the ability to operate safely NPP also in the long-term. In this framework, it is worthy to note that more than 67% of the reactors in operation have to face ageing as they are more than 30 years old. This paper focuses on unsupervised Machine Learning (ML) and artificial neural network (ANN) approach for anomaly detection of SCCs of NPPs. These methods, based on mahalanobis distance and autoencoder neural networks respectively are mainly described including tasks of data analysis, monitoring, prognostics etc. Both ML and ANN were tested on anomaly pattern that deviates from nominal/normal plant conditions. LTO condition is also considered. To the aim of this study, the dataset is provided by a digital twin of primary pipe under inner temperature of 300?C and internal pressure of 15.5 MPa. Finally, the two approach are compared for performance assessment. The findings suggest that the implemented methodology is able to predict pipe failure. The transition from time-based maintenance to predictive maintenance demonstrates to support in a profitable way NPP operation and LTO program allowing also to increase the value of nuclear reactor assets by potentially precluding serious consequences due to faults and failures of plant components.

15.09.2022 08:30 Invited lecture fusion

Fusion – 1000

Overview of the JET Deuterium-Tritium fusion experiments

Henri Weisen

Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne, Switzerland


After more than a decade of preparations, the JET device has in 2021 proceeded to a series of high
power D-T experiments culminating in pulses producing ~10MW of fusion power for a duration of 5
seconds. This power exceeds by a factor 2.5 the fusion power produced in the best similar 5 second
pulse in the previous D-T experimental campaign in 1997. The most important achievements of the
recent JET experiments however reside in the wealth of new physics, most of which remains to be
analysed, harvested in hitherto unexperienced plasma conditions. These include the effects of isotope
composition and energetic ions, such as fusion-produced alpha particles, on plasma confinement.
Important aspects relevant to the operation of a future fusion reactor are also addressed, such as
tritium retention in the plasma facing components and post-experiment tritium clean-up. The JET
results are invaluable for helping to pave the way for tackling the challenges ahead. These include
scaling to reactor size of pulse duration, heat and helium ash exhaust as well as potentially deleterious
plasma instabilities and their control. These issues are being tackled on a broad front by the many
individuals and laboratories participating in the global fusion research effort with the aim of preparing
for and supporting the operation of the international ITER device and designing a future
demonstration power plant.

15.09.2022 09:10 Fusion

Fusion – 1001

Neutron Irradiation Effects in Tungsten for Fusion Applications

Dmitry Terentyev

SCK CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 MOL, Belgium


DEMO and fusion power plants beyond it require robust materials to ensure durable and safe operation as well as commercially competitive construction and dismantling design. One of the main challenges in the development of those materials is assessment of irradiation effects, originating from the nuclear fusion reaction which generates 14 MeV neutrons damaging the material’s atomic lattice. So called in-vessel materials will experience the most severe neutron exposure being far beyond the damage limits acquired by currently operating nuclear power plants. The task of development and qualification of the in-vessel materials thus boils down to securing that the degradation of mechanical, thermal and physical properties will remain within acceptable limits, which are in turn driven by the design of components and operational scenario. This contribution reviews the results of recent irradiation campaigns applied to tungsten – which is the baseline armour material for ITER and DEMO plasma-facing components [1,2].
Driven by the technological priorities, the irradiation tests campaigns were arranged material at the test reactor BR2 (Belgium) in three waves. The first one involved baseline materials focusing on delivery of the engineering design data, the second one targeted screening irradiation of the advanced materials which already passed numerous non-irradiation testing, and the third one included innovative materials with low TRL but essential prospective for applications beyond ITER. Extraction of the properties of the neutron exposed materials involved extensive post irradiation examination (PIE) campaigns in Belgium and Germany to deliver thorough information on the performance of the neutron exposed materials in terms of: mechanical, microstructural, chemical and physical properties. The most important recent results signifying our current understanding of the irradiation effects are reported in this contribution. The performance of the advanced materials is also assessed and presented, which already at this stage allows drawing some important conclusions.
The contribution consists in four chapters focusing on (i) comparative assessment of several commercial tungsten grades complying with ITER specifications; (ii) fracture mechanics testing and analysis to deduce ductile-brittle transition shift in baseline tungsten; (iii) in-depth transmission electron microscopy analysis revealing structural and chemical modifications; and (iv) recent progress in the development of radiation tolerant tungsten-based composites. The talk is concluded with the overview of the current status of material property handbook for tungsten, analysis of the existing gaps and design needs, thereby encompassing future studies.
[1] G. Pintsuk, Fusion Engineering and Design 146 (2019) 1300–1307;
[2] EUROfusion project funded by the Euratom research and training programme 2014–2018 and 2019-2020 under grant agreement No 633053.

15.09.2022 09:30 Fusion

Fusion – 1002

Electrolytic Deuterium Implantation in Metallic Wires

Yannick Verbelen

Polymer Group, H.H. Wills Physics Lab. University of Bristol, Tyndall Ave., Bristol BS8 1TL, United Kingdom


Deuterium (H-2) is one of two hydrogen isotopes of interest for nuclear fusion applications, the other being tritium (H-3). Extensive research has been done on nuclear fusion conditions in a plasma containing H-2 or H-3, however, in electrostatically confined fusors, there are engineering limits to the gas pressure in the chamber. This in turn restricts the concentration of either hydrogen isotope as fuel for fusion, limiting output flux when IEC fusors are used as a neutron source. One approach to improve the probability of nuclear reactions occurring is to use a highly concentrated target, in practice a solid state target rather than a gaseous one in a traditional Hirsch-Farnsworth fusor. In this work, the design and development of an automated parametric apparatus for electrolytic implantation of deuterium in metal wires is presented. Through carefully chose geometric symmetry of the electrodes, a homogeneous implantation of deuterium ions in metal surfaces can be obtained. The implantation speed and depth can be controlled through an electronic interface. The proposed implementation is tested with a zirconium wire target, a popular choice because of the low neutron cross section of Zr. Deuterium implanted in the zirconium is analysed by sampling coupons of the wire after treatment, and implantation concentrations as well as implantation depth are verified using SIMS analysis. The SIMS results are then used to build a parametric model of the implantation system and set the software configuration.

15.09.2022 09:50 Fusion

Fusion – 1003

Cross-section analysis for neutron producing fusion reactions in pre-fusion power operation of the ITER tokamak

Andrej Žohar1, Anders Hjalmarsson2, Žiga Štancar1, A. Trkov3, Luka Snoj1

1Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

2Uppsala University Department of Physics and Astronomy EURATOM-VR Ass., Box 516, SE-75120 Uppsala, Sweden

3Institut Jožef Stefan, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Fusion reactors, such as the ITER tokamak, begin their operation with plasmas that do not emit neutrons, such as the hydrogen plasma. As the first wall is covered with beryllium, plasma contains some beryllium as an impurity. Reactions between plasma fuel ions with beryllium can produce neutrons. The main neutron emitting reactions in the initial phase of tokamak operation are fusion reaction between hydrogen and beryllium ions (9-Be(h,n?)9-B) and the reaction between helium-3 and beryllium ions (9-Be(3-He,n?)11-C).
Neutron production in early fusion non-neutron plasmas is small compared to later power operation in a deuterium-deuterium or deuterium-tritium plasmas. Nevertheless, neutrons emitted from such non-neutron fusion plasmas provide the opportunity to commission diagnostic systems before power operation begins.
To computationally support experimental measurements and the commissioning of diagnostic equipment, reaction cross sections are crucial for the calculation of neutron emission rates. In contrast to the major fusion reactions, such as the deuterium-deuterium and the deuterium-tritium reactions, for which the reaction cross sections are well known and available in evaluated nuclear data libraries, the reaction cross sections for plasma ion reactions involving beryllium are less well known and for helium-3 reactions are not available at all in the evaluated nuclear data libraries.
In this paper, the cross sections for the reaction between plasma fuel ions and beryllium will be analyzed based on measured of cross sections collected in the EXFOR data base and physical models describing the cross sections for interactions between ions. In addition, available cross sections from all evaluated nuclear data libraries, such as the ENDF/B- VIII.0, will also be analyzed. The focus of the study will be on the reactions expected to be produced in the hydrogen and helium plasma of pre-fusion power operation of the ITER tokamak.

15.09.2022 10:50 Severe accidents and PSA

Severe accidents and PSA – 1101


Alberto García-Herranz1, César Queral2, Sergio Courtin1, Marcos Cabezas1, Julia Herrero-Otero1, Enrique Meléndez Asensio3, Miguel Sánchez Perea3, Rafael Mendizábal4

1Universidad Politécnica de Madrid, Alenza 4, 28003, Madrid, Spain

2Universidad Politécnica de Madrid (UPM), Nuclear Engineering Department, José Gutiérrez Abascal 2, 28006 Madrid, Spain

3Nuclear Safety Council, C/Justo Dorado 11, 28040 MADRID, Spain

4Consejo de Seguridad Nuclear, Calle de Pedro Justo Dorado Dellmans, 11, 28040 – Madrid, Spain


After the Fukushima accident, Nuclear Power Plants (NPPs) have had to demonstrate their reliability by increasing their safety provisions and equipment to mitigate accidents. These requirements must cover equipment, procedure improvements, and regulatory development. One of the most important actions has been the introduction of the analysis of the Design Extension Conditions (Multiple Failure Conditions outside the design basis envelope, DEC) into the design requirements of NPPs to further improve safety.

On this aspect, in the framework of the SPAR-CSN project to create a Standardized Plant Analysis Risk (SPAR) model for Spanish NPP, a novel methodology was developed to identify sequences that are outside the design basis but not reaching fuel damage, DEC-A, making use of the PSA level 1 model for a generic plant (3-loop PWR Westinghouse design). Using this methodology, we have been able to identify the DEC-A sequences with the highest relative risk by comparing their frequencies with their corresponding DEC-B (sequences with fuel damage).

As it is shown in this study, the addition of Post-Fukushima improvements reduces the relative risk of the DEC sequences in which these are involved.

15.09.2022 11:10 Severe accidents and PSA

Severe accidents and PSA – 1102

Numerical study of melt penetration into a particulate bed

Liang Chen, Walter Villanueva

Royal Institute of Techology, Div. Of Nuclear Power, Brinellv. 60, S-10044 Stockholm, Sweden


In a severe accident scenario of a light water reactor, molten corium can fall into the lower head of the reactor pressure vessel (RPV) and form a particulate debris bed consisting of oxidic and metallic particles. Due to decay and oxidation heat, the lower-melting metallic particles of the debris can re-melt and then penetrate mainly downwards the porous debris bed and possibly relocate to the bottom of the RPV. In this case, the configuration and properties of the debris bed change and affect the mode and timing of possible failure of the RPV especially in the reactor designs with vessel penetrations at the bottom. In this study, we investigate the effect of the wettability of the non-isotropic particulate bed on melt penetration kinetics. A two-phase, non-isothermal, laminar, and incompressible Newtonian fluid flow is set-up to model and simulate the heat and mass transport of molten metal in a fixed matrix of particle bed. A complementary set of experiments using 1.5 mm particles and Sn-Bi melt are also performed to validate the models. The ultimate goal is to generate correlations that can be implemented in integral Severe Accident (SA) codes.

15.09.2022 11:30 Severe accidents and PSA

Severe accidents and PSA – 1103

Pool Fire Accident Simulation Validation Assessment

Nima Fathi1, Jared Kirsch2

1Texas A&M University, Marine Engineering Tech Department, P.O. Box 1675, Galveston, TX 77553-1675, USA

2National Renewable Energy Laboratory, 1617 N. Cole Boulevard, GOLDEN, COLORADO 80401, USA-Colorado


Pool fire accident scenarios are potentially highly dangerous phenomena which can occur when hydrocarbon fuel is spilled in a medium including nuclear materials. Understanding the multiphysics phenomena exhibited by such fires and improving pool fire modeling tools are important steps toward enhancing safety in these scenarios. In this study, a 30-cm diameter methanol pool fire was modeled using Sandia National Laboratories SIERRA/Fuego turbulent reacting flow code. Large Eddy Simulation (LES) with subgrid turbulent kinetic energy closure was used as the turbulence model. Combustion was modeled and simulated using a strained laminar flamelet library approach. Radiative heat transfer was modeled using the gray-gas approximation. Validation analysis was conducted on the resulting dataset. The area validation metric (AVM) was used to quantify simulation uncertainty and to evaluate the performance of the model. Solution verification including the grid convergence index was performed, and the multivariate validation metric was applied to temperature and axial velocity as the main system response quantities in our analysis. This study aims to expand the validation analysis of the pool fire model and thus provide insight into the model’s performance and possible future improvements.

15.09.2022 11:50 Severe accidents and PSA

Severe accidents and PSA – 1104

Experiments on radiolysis gas detonation in BWR exhaust pipes and mechanical response to the detonation pressure loads

Mike Kuznetsov1, Joachim Grune2, Reinhard Redlinger2, Wolfgang Breitung1

1Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany

2Karlsruhe Institute of Technology (KIT), Hermann von Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen, Germany


Radiolysis gas (2H2+O2) can accumulate in BWR steam piping in case of steam condensation, with an ensuing detonation of the radiolysis gas being the likeliest cause of a pipe and/or valve rupture. In the current work we consider a typical BWR exhaust pipe, which connects the high pressure steam piping with the ambient atmosphere, under the following ”worst case” scenario: (a) accumulation of radiolysis gas in an exhaust pipe, (b) fast valve opening to the high pressure system with steam at 70 bar, and (c) adiabatic pressurization of the radiolysis gas by the steam. The main purpose of the current work is an experimental and numerical evaluation of the maximum pressure load plus the integrity of the BWR exhaust pipe in case of a detonation of the pressurized radiolysis gas. Detonation experiments of radiolysis gas were performed in a real scale exhaust pipe of 12.25 m length made from stainless steel Nr. 1.4541. To reproduce the ”worst case” conditions for radiolysis gas after the steam pressurization we used a radiolysis gas mixture at 10 bar and 293 K which energetically (with respect to the pressure load) equals the pressurized BWR mixture at 20 bar and 602 K. The inert part of the tube was filled with nitrogen instead of steam to enable experiments at ambient temperature. The radiolysis gas was included in a thin shell vessel to separate it from the inert gas, with ignition at a metal plate side used as an imitation of the gas-water interface in real conditions. 16 strain gauges were installed along the tube to measure longitudinal and circumferential deformations under the detonation process, and dynamic pressure loads were measured using 4 pressure transducers installed on flanges. Four experiments were performed: at 1.6, 5.0 and two tests at 10.0 bar. The experiments showed that maximum deformations occurred at the end of the radiolysis gas vessel. The maximum dynamic strain was measured to be 0.75% for the radiolysis gas detonations at 10 bar, with the maximum remaining deformation, using strain gauges and direct measurements, being about 0.15%. This means that the exhaust tube remains intact even under this worst case detonation scenario of the pressurized radiolysis gas mixture.
To simulate detonation of the radiolysis gas mixture at 20 bar and 602 K with steam as an inert gas, the 3-dimensional CFD code DET3D [1] was used. Using a simplified 1D model for the mechanical response of a cylindrical pipe under an internal dynamic pressure load, the dynamic strain corresponding to these calculated pressure signals was determined. A comparison of these calculated strain values with the experimental signals showed a very good agreement. This made it possible to use the DET3D code to predict the detonation pressure loads for the original BWR gas mixture under the worst case scenario, and to evaluate the resulting dynamic deformations and the integrity of the exhaust pipe.

15.09.2022 12:10 Severe accidents and PSA

Severe accidents and PSA – 1105

Influence of non-condensable gas-dust mixture on direct contact condensation of steam at atmospheric pressure

Luca Berti, Alessio Pesetti, Guglielmo Giambartolomei, Donato Aquaro

University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy


At the Department of Civil and Industrial Engineering (DICI) of the University of Pisa, an experimental research program, funded by the ITER Organization, concerning steam, non-condensable gas and dust injection into the Pressure Suppression Tanks was carried out.
This mixture of fluids and dust is produced in the case of a Loss of Coolant Accident in the ITER Vacuum Vessel. The aim of the research program is to determine the steam condensation efficiency in such conditions.
Experimental tests were performed injecting this mixture in a tank partially full of water. Alumina was used as simulant of the actual dust present in the ITER Vacuum Vessel. During the experimental tests mass flow rates, temperature and pressure of the different fluids involved were recorded.
The experimental rig, built at the “B. Guerrini” Laboratory of the DICI-University of Pisa, consists of a condensation prismatic container system with two holed spargers located inside; a superheated steam supply system; two air tanks pressurized up to 10 bar; a degassed water supply system; a dust micro-dosing system; an air discharge line, containing a demister and HEPA filter; a data acquisition and control system and a visualization and video recording system. The prismatic container (w=1460 mm, L=2300 mm and H=990 =mm), insulated by rock wool, is filled with 2.230 m3 of water.
The experimental tests were performed injecting dust (3/6/12 kg, 0.67-1.11 g/s) mixed with a steam mass flow rate (11.49-25.02 g/s, T=150 °C, 1.5 bar) and a primary air mass flow rate (5.71-7.22 g/s, Tap=67.03-71.82 °C) in the water container (Tw=22.7-100 °C, P=1.013 bar, water head=0.665 m). After the injection of the total mass of dust, steam is injected into the water tank until it reaches the temperature of 100 °C.
The steam condensation into the subcooled water pool was investigated to characterise the condensation regimes occurring during the mixture injection. The condensation efficiency depends on the steam jet length and on the effective heat transfer coefficient. Few grams of dust reduce the water transparency, to overcome this drawback preliminary separated tests without dust were carried out. The influence of dust and non-condensable gas was determined comparing the pool temperature increase, determined by equal steam mass, in the tests with steam alone.
Measurements of the lengths and surfaces of the mixture jets were performed by means of image elaboration and were compared with theoretical correlations. The comparison between steam injection and steam-air injection at water temperature of 30 °C showed a strong influence of the air above all on the jet shape and on the heat transfer coefficient.