Preliminary Program of the 29th International Conference NENE2020

08.09.2020 15:40 Poster Session

Reactor physics - 201

On the Self-Shielding in the Unresolved Resonance Range

Andrej Trkov

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


Nuclear reaction cross section have energy-dependent behavior, where the cross section can change by several orders of magnitude over narrow energy intervals – these are called resonances. The relative resonance density increases with energy to the point, where they cannot be resolved experimentally, but the fluctuations in the cross sections still contribute to the so-called self-shielding effect. Self-shielding occurs because resonances deplete neutrons near the resonance energies, thus effectively reducing reaction rates. Standard text-books on reactor physics address the self-shielding effect in the unresolved resonance range using various approximations by defining average resonance parameters, which are then used to derive self-shielded cross sections. An alternative method usually applied in Monte Carlo transport codes is to define probability tables or multi-band parameters to tackle self-shielding.

An exercise was performed at the International Atomic Energy Agency (IAEA) to validate methods of processing nuclear data to make application libraries in the so-called ACE format for use in various Monte Carlo transport codes (“”). Probability tables give the probabilities that a cross section in a certain energy interval lies within a certain range of cross section values. About 20 bins are commonly used in the libraries in ACE format available from the Los Alamos National Laboratory (LANL). Multi-band parameters are derived by the conservation of moments. Two bands are usually sufficient. Previous exercise has demonstrated that multi-band parameters and probability tables can be used interchangeably (“”).
In an effort to provide to the IAEA Member States an independent route to generating application libraries, a project was initiated that produced the ACEMAKER code [ref.Daniel], which is a module for assembling partly processed data using the PREPRO package (“”). The GROUPIE module of package version PREPRO2019 has been reorganized to simplify the processing.

Details of the method for generating the multi-band parameters are described. The method does not rely on the average resonance parameters, but extrapolates moments directly from the resolved resonance range. The assembled library files in ACE format are tested against a suite of criticality benchmark experiments. Results using ACE libraries generated by different data processing codes (starting from the same evaluated nuclear data files for 235U and 238U) are shown.

08.09.2020 15:40 Poster Session

Reactor physics - 202

Extension of RAPID for TRIGA Reactor Real-Time 3D Burnup Calculations

Anze Pungercic1, Valerio Mascolino2, 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


Determination of accurate 3D pin-wise fuel burnup in nuclear reactors is vital from the standpoint of fuel management, spent fuel storage safety and safeguards. In addition, the need for accurate and efficient burnup calculations has become more urgent for the simulation of advanced reactors and monitoring of spent fuel pools. To accomplish this, the Virginia Tech Transport Theory Group (VT3G) has been working on advanced computational tools for accurate modeling and simulation of nuclear systems in real-time. One such capability is a novel methodology for performing 3D fuel burnup calculations, bRAPID, which utilizes the RAPID Code System. RAPID is based on the Multi-stage Response-function Transport (MRT) methodology, that decouples a problem into independent stages that are then coupled in real-time via transfer functions/coefficients.
Recently, we initiated activities to benchmark the bRAPID methodology using the well characterized Jozef Stefan Institute’s TRIGA Mark-II research reactor. Thus far, we have created a database including full operational history that allows for burnup validation possibilities in the form of measured excess reactivity. Additionally, for further confirmation, we are planning to perform burnup measurements using the fuel gamma spectrometry.
In this paper, extension of the bRAPID algorithm for its application to the TRIGA research reactor will be presented. In particular, the paper will focus on bRAPID’s database pre-calculation procedure and its automation. The behavior of the Fission Matrix (FM) coefficients for different combinations of reactor power and irradiation times will be analyzed. Finally, the methodology will be compared to already validated Serpent-2 burnup calculations for the latest JSI TRIGA core configurations where changes in keff and isotopic composition will be presented.

08.09.2020 15:40 Poster Session

Reactor physics - 203

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 Libraries for the Nuclear Design Calculations of the NPP Krško with the CORD-2 System

Marjan Kromar1, Bojan Kurinčič2

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

2Krško Nuclear Power Plant, Vrbina 12, 8270 Krško, Slovenia


Recently two new nuclear data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since their release, many researchers have been investigating how the existing calculation results in a given system are influenced by the new evaluations. The purpose of this study is to examine the effects of the newly cross sections libraries on the nuclear design calculations of the NPP Krško core.

In the nuclear design process two different types of calculations are performed:

1. calculation of the neutron transport in the media, where neutron transport (or diffusion) equation is solved to obtain spatial neutron flux distribution and system reactivity,
2. fuel depletion, where Bateman equations are solved to obtain the time evolution of nuclide concentrations.

Libraries have profound impact on both aspects. In this paper CORD-2 system is used for the library comparison. In CORD-2, depletion and lattice cell calculations are done with the well-known WIMS-D5 code, while the 3-D core calculations are performed with the GNOMER diffusion code. In the first part of the paper the effect on the depletion of the typical NPP Krško fuel assembly in infinite geometry is investigated. In the second part, analysis of all 30 completed NPP Krško operating cycles is performed. Comparison of the results gives some indications of the libraries performance, while comparison to the measurements performed on the plant gives some clues how to improve CORD-2 nuclear design capabilities.

08.09.2020 15:40 Poster Session

Reactor physics - 204

Validation of Serpent-PARCS code sequence using fuel cycle 1 of KRŠKO NPP

Dušan Čalič1, Marjan Kromar2

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

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


At the Jožef Stefan Institute the CORD-2 simulator is used for core design calculations of NPP Krško since 1990. In 2016 new approach was developed where the existing WIMSD-5B code for lattice cell calculations was replaced with Monte Carlo code Serpent 2. The effective diffusion homogenization method (EDH) was used to obtain the cell homogenized cross using Serpent code and to use them in fuel assembly calculations. The results of radial power distribution and critical boron concentration turned out to be in good agreement at HZP and HFP conditions for cycle 1 however the study demonstrated that the use of three step procedure using Monte Carlo code is (still) out of the question for routine calculations due to large computational time. For this paper the results of two step procedure using Serpent and PARCS code will be presented and compared to existent results of cycle 1.

08.09.2020 15:40 Poster Session

Reactor physics - 205

Evaluation of cross section and fission yields induced uncertainty in the VVER-440 burnup calculation

Branislav Vrban, Stefan Cerba, Jakub Lüley, Filip Osuský, Vladimir Nečas

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


The properties of nuclear fuel depend on the actual isotopic composition which develops during a reactor operation. In practice, the prediction accuracy of burnup calculations serves as the basis for the future precise estimation of a core lifetime and other safety-based core characteristics. The present study quantifies nuclear data induced uncertainties of nuclide concentrations and multiplication factors in VVER-440 fuel depletion analysis. The well-known SCALE system and the TRITON sequence are used with the NEWT deterministic solver in the SAMPLER module that implements stochastic techniques to assess the uncertainty in computed results. The propagation of uncertainties in neutron cross section and fission yields is studied through the depletion calculation of 2D heterogenous VVER-440 fuel assembly with an average enrichment of 4.87 wt % 235U and six gadolinium rods with 3.35 % of Gd2O3. In the paper, fixed nominal depletion conditions are based on the real operational data of Slovak NPP Bohunice unit 4 during cycle 30. In total 250 cases with uncertain parameters are computed and the results are evaluated by an auxiliary tool.

08.09.2020 15:40 Poster Session

Reactor physics - 206

Effect of Nuclear Data Libraries on PWR Ex-core Detector Response

Tanja Goričanec1, Andrej Trkov2, Klemen Ambrožič2, 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


Knowing the reactor power is of utmost importance for safe operation of a nuclear power plant. In a typical pressurized water reactor during normal operation, reactor power is monitored by power-range detectors positioned outside the reactor core. With the aim to predict their response, a detailed core and ex-core model using Monte Carlo Neutron Transport Code MCNP were developed. Due to the ex-core position of neutron detectors, hybrid code ADVANTG is used to generate weight windows to speed up neutron transport outside the reactor core. To be able to use ADVANTG, fixed neutron source had to be reconstructed from the criticality core calculation. Different fixed source descriptions were compared in previous research [1], where the need for describing pin wise neutron source was identified. This paper focuses on the evaluation of the effect of different nuclear data libraries on the ex-core detector response. The first step is criticality core calculation aiming to calculate fixed neutron source for the ex-core calculation. Comparison of reactor power distribution obtained in the criticality core calculation using ENDF/B-VII.0 and ENDF/B-VIII.0 nuclear data libraries was performed. The deviation between both libraries was below 1 % and was considered insignificant. The results from the criticality core calculation were used to produce fixed source description for the ex-core calculation. The next step was the evaluation of the fixed source generated using different versions of the ENDF/B nuclear data libraries, where the focus was on the description of the prompt fission neutron spectrum. Different methods for the fixed source spectrum description were studied. The first method involved calculating prompt fission neutron spectrum by sampling fission neutron generation rate within the core, fission neutron energies and positions. In this method, fission neutrons were divided into energy bins, which were used to describe bin-wise neutron spectrum for fixed source calculation. In the second method, Watt fission spectrum was fitted to energy distribution from the first method. First and second method were investigated with a model taking into account only the core average spectrum and a model covering 24 axial layers of individual fuel assemblies. The third method for describing prompt fission spectrum included combining prompt fission spectra of 235U, 239Pu and 238U, weighted by the calculated fission reaction rates in 24 axial layers for individual fuel assemblies. In the last part of the work, neutron transport outside the reactor core using different versions of ENDF/B-VII.0 nuclear data libraries was studied to evaluate ex-core detector response. The deviation between the two nuclear data libraries (used for fixed source description and ex-core neutron transport) was ~7 %: the major reason for the large difference is isotope 56Fe, which is deficient in the ENDF/B-VIII.0 evaluation. When replacing the cross section evaluation for 56Fe with the improved one from the IAEA INDEN project, the deviations between both libraries due to the 56Fe decreased to ~2 %. The analysis showed that correct description of the neutron spectrum is very important for ex-core calculations, especially in the fast neutron energy region.

[1] T. Goričanec, et al., “Sensitivity analysis of ex-core detector response using different neutron sources in a typical PWR using Monte Carlo methods”, International Conference on Radiation Shielding 2020, Seattle, WA, USA, September 2020.

08.09.2020 15:40 Poster Session

Reactor physics - 207

Methodology for Thermal Neutron Scattering Cross Sections Determination

Ingrid Vavtar, Andrej Trkov, Luka Snoj

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


Knowledge of thermal nuclear data is applicable in many scientific fields and is heavily interdisciplinary. Using neutrons for research enables us to investigate the world around us, as well as to develop new materials and processes to meet the needs of the society. In the near future local electricity needs will greatly increase due to the increasing adoption of electric cars, air-conditioning units, heat pumps, as well as growing and urbanizing global population. In order to generate carbon-free electricity, radical innovations in renewable energy are required. Due to fluctuations in electricity generation from solar and wind sources, an additional reliable sources of electricity that can track and compensate for these fluctuations quickly enough is needed. Small modular reactors in addition to reliable electricity production capabilities can provide also fast response to fluctuations in electricity production due to prompt negative temperature reactivity coefficient. The main contributions to the prompt negative temperature reactivity coefficient are the Doppler effect and thermal spectrum hardening. Accurate prediction of these effects is highly dependent on precise calculations or simulations of the reactor physical parameters, where knowledge of nuclear data and their uncertainties at thermal energies are needed.
Currently used thermal scattering nuclear data are based on old evaluations or evaluations which have been improved only slightly over the past few decades. These old data sets do not include any uncertainty information. For the purpose of following the production of electricity from renewable sources, new sets of nuclear data will be needed, which in addition to the data themselves will include estimated uncertainties. In the low neutron energy range, typically below 5 eV, neutron scattering is affected by the atomic bonding of the scattering molecule in the moderator. Compared to a free nucleus, this changes the reaction cross section and, thus, the energy and angular distribution of the secondary neutrons. The thermal neutron scattering cross sections need to be generated by employing state-of-the-art atomistic simulations which will be presented in the article. Methodology for thermal neutron scattering cross sections determination involves use of three different programs, such as density functional theory capable computer codes (VASP or Quantum Espresso), programs for lattice dynamics calculations (PHONON or Phonopy) and program for calculation of scattering law from density of states (LEAPR). In the present work the planned strategy for generating thermal scattering data will be outlined.

08.09.2020 15:40 Poster Session

Reactor physics - 208

Nuclear reactions for epithermal neutron dosimetry

Vladimir Radulović1, Andrej Trkov1, Luka Snoj1, Anze Pungercic1, Thiollay Nicolas2, Christophe Destouches3, Hubert Carcreff2

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

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

3Commissariat a l'Energie Atomique - Centre d'Etudes de Cadarache DRN/DER, Bat 238, 13108 St Paul Lez Durance Cédex, France


Reactor dosimetry is the primary method of determination of neutron fluxes and fluences in neutron fields such as in fission and fusion nuclear reactors, neutron generators. Threshold activation reactions, e.g. (n,p) reactions and (n,?) reactions, etc., are generally used for characterizing fast neutrons with an energy threshold around or above 1 MeV, while radiative capture reactions (n,?) are dedicated to neutrons of thermal and resonance energies (up to a tenth of eV). Few nuclear reactions are sensitive specifically to neutrons in the intermediate (epithermal) energy region which are of primary importance for applications like 4th generation sodium-cooled fast reactor as well as in the structures of fusion device (JET, WEST and ITER).
Previous works have been performed by the CEA for identifying new reactions and dosimeters leading to the testing of Zirconium dosimeters (92Zr, 94Zr capture reactions under Boron neutron filter) and Tin (117Sn inelastic reaction) dosimeters. JSI and CEA have also recently studied thermal and low epithermal energy reactions through irradiations performed in the JSI TRIGA reactor and using adapted neutron filters. These studies stated the feasibility of the use of these reactions for characterization of the epithermal spectrum and improved knowledge on the use of neutron filters. They also revealed the need to enhance associated nuclear data (high cross section uncertainties mainly) and the experimental method limitations (perturbations induced by neutron filters).
As the 250 kW JSI TRIGA reactor offers an efficient neutron calculation scheme and very well characterized irradiation locations in terms of the knowledge of the neutron and gamma fields, thanks to the work performed at the JSI over the last decade mostly in collaboration with the CEA, conditions are gathered to deepen the common research on epithermal dosimeters.
After a reminder of the state of the art on the existing means for characterization of epithermal neutron spectrum, this paper summarizes the main previous results obtained for epithermal dosimeters (Zirconium, Tin) and neutron filters. Their analysis leads to the definition of a research program of which the principle is described hereafter. First, realisation of an extended review of possible inelastic reactions for identification of new candidate reactions. Second, setup of the principle for deriving epithermal fluxes and nuclear data from experimental results obtained with or without neutron filters to correct thermal contributions. Specific attention will be paid to the analysis of format of the experimental results in order to provide valuable data for nuclear data evaluators. Third, definition of the mains experimental features of the foreseen irradiation campaign for testing new dosimeters and for upgrading data for Zirconium and tin dosimeters (irradiation locations, irradiation device, experimental process for irradiation and activity measurements). It will rely on a specific neutron modelling of the irradiation conditions. Eventually, expected timeline of the programme will be given.

08.09.2020 15:40 Poster Session

Reactor physics - 209

Experimental validation of RAPID based on JSI TRIGA reactor dosimetry

Valerio Mascolino1, Vladimir Radulović2, Alireza Haghighat1, Luka Snoj2

1Virginia Tech Northern Virginia Center Mechanical Engineering Department, Nuclear Engineering Program, 7054 Haycock Rd, Falls Church, VA 22043, USA-Virginia

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


The RAPID (Real-time Analysis Particle-transport and In-situ Detection) Code System, based on the Multi-stage Response-function Transport (MRT) methodology, utilizes the Fission Matrix (FM) and Detector Response Function (DRF) approaches for calculation of the fission neutron source and detector responses / doses respectively. The MRT methodology decouples the analysis of a nuclear system in several physics-based stages, for which response functions or coefficients are pre-calculated as a function of problem-relevant parameters (e.g., control rod positions for TRIGA reactors). These response functions are then coupled in real-time via linear systems of equations for any input configuration.

As part of an ongoing collaboration between Virginia Tech and the Jožef Stefan Institute (JSI), the code is undergoing experimental validation and benchmarking using the JSI TRIGA Mark-II reactor. RAPID has already been validated at critical condition for different core configurations. In this work, RAPID is being validated for dosimetry calculations using in-core foils of Aluminum (99.9% wt.) – Gold (0.1% wt.), and ex-core 27Al(n,?) and 197Au(n,?) dosimeters. The paper will compare RAPID results to both measurements and a standard 3-D Serpent Monte Carlo prediction.

The aim of this work is to demonstrate the ability of RAPID of accurately calculating doses (and, in general, reaction rates) in real-time (seconds), making the code a useful tool, e.g. for dosimetry calculations, pressure vessel fluence calculations, and design of experiments in a fraction of the time required using state-of-the-art codes.

08.09.2020 15:40 Poster Session

Reactor physics - 210

3D Reactor Core Modeling

Luka Štrubelj1, Klemen Debelak1, Dušan Čalič2

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

2ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia


First a 2 step (fuel element, core) calculations are demonstrated with Serpent-PARCS coupling codes for various fuel elements by studying the homogenization procedure on a fuel assembly level using Serpent code. The results were compared with exact full scope Monte Carlo results, where the average difference of power distribution was 1.7 % for specific core loading pattern.
Based on a lesson learned from the Serpent homogenization procedure a 3D model of nuclear reactor core and quasi 3D model of reactor pressure vessel was developed in computer program Apros. The goal is to perform transient reactor kinetic calculations with reactor coolant response. First the 2D homogenization of cross-sections were calculated in Monte Carlo code Serpent. The calculation is done for different types of fuel in terms of enrichment, burnable poisons and burnup steps. The corrections (reactivity coefficients) are calculated for changes in boron, fuel temperature and moderator temperature (density) change. Reflector is modelled with albedo coefficients. The cross-sections are transferred to appropriate format as input for Apros. The steady state simulation at full power is run in Apros. Two analyses are performed: symmetrical control rod bank insertion and asymmetrical insertion of one control rod. The 3D results are analysed.

08.09.2020 15:40 Poster Session

Research reactors - 301

On the optimisation of large sample in-core irradiation channel in the JSI TRIGA reactor

Tanja Goričanec1, Sebastjan Rupnik2, Anže Jazbec2, Luka Snoj3

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

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

3Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


One important feature of the Jožef Stefan Institute TRIGA Mark II reactor is the so called triangular channel in the reactor core. The channel occupies three irradiation positions and similar neutron flux (1.0×1013n/cm2s) as the central channel (1.9×1013 n/cm2s), but has more than three times (37 cm2) larger surface area. Thus making it suitable for irradiation of large samples in the core. It is mostly used for radiation hardness studies of detectors and electronic equipment. Due to high demand for irradiation in triangular channel, a decision was made to introduce another triangular channel into the core. Hence it was decided to make two of them, one new one and another one to replace the old one. The aim of this research was to optimize new channel design, to enable optimal experimental conditions and to keep operational limits and conditions within the specifications. A detailed computational design for different possible geometrical configurations of new triangular channel using Monte Carlo neutron transport code MCNP were made. Four different configurations were studied: triangular channel reaching to the bottom grid plate without any insert (first configuration), triangular channel reaching to the bottom grid plate with graphite (second configuration) and aluminium (third configuration) insert and lifted triangular channel to approximately middle of the active core height without any inserts (fourth configuration). All configurations were also compared to the calculation without modelled triangular channel (bare configuration), where three fuel elements were in position of triangular channel. It was found out that the impact on multiplication factor is from -730 pcm to -1020 pcm, depending on configuration studied. The hot rod power peaking factors was calculated to be from 1.393 to 1.403. The relative deviation in total neutron flux in central channel, compared to the bare configuration, was evaluated to be from -1.45 % to -1.72 %. The deviation in total neutron flux in triangular channel between different triangular channel configurations was established to be up to 6 %. The total power density peaking factor, axial power peaking factor and radial power density peaking factor were calculated to be approximately 2.70, 1.27 and 2.09 respectively, and did not deviate significantly between different configurations. It was confirmed that triangular channel has negligible effect on neutron spectrum in central channel studied in 640 energy groups. When studying 640 group energy spectrum in different triangular channel configurations it was found out that the highest thermal peak has configuration 4, followed by configuration 2, which is expected due to the presence of inserts. The study of rotary groove showed high impacts of triangular channel on neutron flux in 3 energy groups, however the deviations between different channel designs were not significant. Radiation hardness factor and 1 MeV flux equivalent by ASTME standard deviated between different configurations up to 3.5 %. To sum up, calculations confirmed no significant deviation between different channel designs and final design can be chosen based on technical limitations.

08.09.2020 15:40 Poster Session

Research reactors - 302

Water Activation Experiments and Calculations at JSI TRIGA Research Reactor

Andrej Žohar1, Vladimir Radulović1, Anže Jazbec2, Igor Lengar1, Sebastjan Rupnik2, Luka Snoj1

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

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


Demineralized water is the cooling fluid in the majority of fission nuclear power plants, research reactors and it is also considered as a cooling fluid in fusion reactors. As water is exposed to neutrons it becomes activated. There are several activation reactions of which the most important is the O-16(n,p)N-16 reaction due to high natural abundance of O-16 and the high energy of emitted gamma rays (~ 6.13 MeV and 7.11 MeV) with 7.13 s half-life. In case of fission reactors the water contribution to the total radiation field is usually negligible. The situation is different for fusion devices such as ITER, where the dose-rate due to activated water is estimated to be several 100 Gy/h thus potentially causing increased doses to personnel, radiation damage to components around cooling systems and additional nuclear heating to superconducting coils.
Fission research reactors present an opportunity to study the activation of water and effects of activated water decay as currently no fusion reactors fusing deuterium and tritium are capable of performing water activation experiments with sufficient accuracy. At the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor, water activation experiments are regularly performed for education of nuclear engineering students.
The water activation system at the TRIGA reactor consists of a water pump and tube taking water underneath the reactor core and transporting it through reactor core to the reactor platform where several detectors (High Purity Germanium, LaBr, etc.) are located for measurements of activated water spectra. The activated water is returned into the reactor pool through the tube.
In this paper recent experimental results with the water activation system at JSI TRIGA reactor will be compared to calculated results with in-house developed methodology. The methodology consists of two steps. The first step is the calculation of water activation in the reactor core using the Monte Carlo program Monte Carlo N-Particle (MCNP) which has been validated on several experiments performed at the JSI TRIGA reactor. The second step of the methodology is the calculation of gamma ray spectra and dose at desired position using the MCNP program and parameters calculated using the first step of methodology.

08.09.2020 15:40 Poster Session

Research reactors - 303

Updated spectral parameters and neutron fluxes in specific irradiation channels in the carousel of TRIGA MARK IPR-R1 research reactor used to apply neutron activation analysis, k0-method

Radojko Jaćimović1, Maria Angela de Barros Correia Menezes2

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

2Nuclear Technology Development Center/Brazilian Commission for Nuclear Energy, Avenida Presidente Antônio Carlos, 6.627, 31270-901 Belo Horizonte, Minas Gerais, Brazil


The TRIGA MARK I IPR-R1 research reactor has been operating since November 6, 1960. It is located at Nuclear Technology Development Centre, CDTN, (Centro de Desenvolvimento da Tecnologia Nuclear) sponsored by Brazilian Commission for Nuclear Energy, CNEN (Comissao Nacional de Energia Nuclear), Belo Horizonte, Brazil. This reactor is licensed and has been operating at 100 kW but its core configuration is ready to operate at 250 kW. IPR-R1 reactor has been used for training of reactor operators, production of radioisotopes (radiopharmaceuticals and radiolabelled molecules to research aiming at future applications in nuclear medicine; radiotracers for studies in analytical chemistry, in environmental and industrial applications; production of radioactive sources for steel industry, etc.) and neutronic and thermohydraulic studies. There are three irradiation devices – central tube, fast pneumatic transfer system and the carousel, however, the carousel is the most used due to some advantages in neutron spectral parameter ratios.
The core configuration has been modified six times since the first criticality and elsewhere there are nine publications about the neutron fluxes determinations from 1975 and 2014, using experimental and semi theoretical methodologies. Related to experimental procedures, several materials/monitors were used and the neutron fluxes were determined in different irradiation channels and devices. Monte Carlo Simulation, MCNP, was applied in semi theoretical procedure. Summarizing, over time, several studies were carried out, determining the neutron fluxes in different irradiation channels and devices, applying different procedures and materials.
From configuration no 5, 1996, thermal and epithermal neutron fluxes were determined and the average values for carousel were published in four papers. These values were calculated experimentally and using MCNP. Until 2001, the carousel was used to rotate and the values determined were the average values for thermal and epithermal neutron fluxes. When the new configuration for 250 kW was made, it was necessary to decide not to rotate the carousel during irradiations in order to preserve the mechanism that was damaged. In 2003, the values were determined experimentally in five specific channels aiming at the application by neutron activation k0-method. The procedure applied was the Cd-multimonitors method, using several certified foils and alloys, and more suitable cadmium boxes. The same procedure was repeated in 2016 and in 2019. Values for thermal, epithermal and fast neutron fluxes as well as f and alpha spectral parameters were determined and the procedure was established. This paper is about these determinations.
This work was partially supported by Brazilian Foundation for Research Support of Minas Gerais, FAPEMIG, under grants PPM-00143-16 and APQ-00588-18, by Brazilian National Council for Scientific and Technological Development, CNPq, under grants PQ 2016 and PQ 2019 and by financial support from the Slovenian Research Agency (ARRS.

08.09.2020 15:40 Poster Session

Research reactors - 304

Jožef Stefan Institute TRIGA Research Reactor Activities in the Period from September 2019 – August 2020

Anže Jazbec1, Sebastjan Rupnik1, Vladimir Radulović2, Andraž Verdir1, Marko Rosman1, 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


Jožef Stefan Institute (JSI) has been operating a 250 kW TRIGA research reactor since 1966 Safety performance indicators (SPI) have been monitored for over 10 years now. Examples of the monitored SPIs are operating time, number of irradiated samples, doses received by operating staff and activity of radioactive gases released to the environment. In the paper, SPIs for the year 2019 will be presented and analysed. Also comparison among SPIs from previous years will be done. Upon the SPIs analysis, future operation can be improved and safety of the reactor can be increased.
Furthermore, new research work carried out during the years 2019 and 2020 will be presented. Most of the research collaborations from previous years continued like collaboration with CEA, where new type of neutron and gamma detectors were tested. Irradiations in the framework of NATO project continued where SiC is being tested as possible neutron detector. Two research campaigns with Rolls Royce Civil Nuclear SAS were performed.
Within a PhD research work of Klemen Ambrožič a set of unique measurements and calculations was performed. For the first time that a full coupled neutron-electron-gamma calculation methodology dealing with both prompt and delayed emissions is adapted and validated on a research reactor characterized by high dose levels.
In collaboration with the University of Lancaster a remotely operated robot designed to provide gamma and neutron dose fields of unknown areas was tested. The robot can be completely autonomous. It uses different types of detector and provided dose rate maps were very accurate. Using the robot, also low active sources were identified in our reactor hall.
In spring 2020, in the light of SARS CoV-2 epidemic, protective masks were tested for sterilization with gamma rays.
In summer 2020, small amounts of Am-241 will be irradiated in order to determine neutron cross-section more accurate that current available data.
In the field of training, practical courses in the field of experimental reactor physics for students for Faculty of Mathematics and Physics were carried out. In addition, course on Nuclear technology was carried out for future Krško NPP operators. In Autumn 2019 we hosted 3 weeks of a 6-week research reactor training course, held at ATI, Vienna and CTU in Prague. For the first time, a group of students from Uppsala University was hosted on a 3-day Practical Reactor Physics Course. The plan is to repeat the course every year. In November, 24 students from University of Milan took a one-day course on Prompt critical state of nuclear reactor.
In the last 12 months, there were some non-routine maintenance done. From the researching point of view, the most important investment are new triangular channels. Now, the reactor can operate with two triangular channels simultaneously. Additionally, the new ones have larger volume so bigger samples can be irradiated inside the reactor core completely surrounded by fuel. The upgraded pneumatic transfer system was successfully commissioned and windows inside the reactor hall were replaced. The data on control rod position was digitalized. That allows us to read control rod position from our digital archive which is useful for some research applications.
For the first time, TRIGA reactor participated at event called “Noč raziskovalcev” or science night. Visitors were equipped by protective clothing and personal dosimeter so they were able to visit the reactor platform during reactor operation.

08.09.2020 15:40 Poster Session

Research reactors - 305

Feasibility of reactor pulse operation at the JSI TRIGA reactor for nuclear instrumentation detector testing at very high neutron flux levels

Vladimir Radulović1, Igor Lengar1, Loic Barbot2, Grégoire De Izarra3, Manuel Cargnelutti4, Danilo Bisiach4

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

2CEA-Cadarache DEN/DER/SPEX/LDCI, Bâtiment 238 - Piece 10, F13108 Saint-Paul-lez-Durance, France

3CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France

4Instrumentation technologies d.o.o., Solkan, Srebrničev trg 4a, 5250 Solkan, Slovenia


Feasibility of reactor pulse operation at the JSI TRIGA reactor for nuclear instrumentation detector testing at very high neutron flux levels

Vladimir Radulović, Igor Lengar
Jožef Stefan Institute
Jamova 39
SI-1000 Ljubljana, Slovenia,

Loic Barbot, Grégoire De Izarra
CEA, DES, IRESNE, Instrumentation, Sensors and Dosimetry Laboratory
CEA Cadarache
F-13108 St Paul Lez Durance – France,

Manuel Cargnelutti, Danilo Bisiach
Instrumentation Technologies, d.o.o.
Velika pot 22
SI-5250 Solkan, Slovenia,

A vital phase in the development of nuclear instrumentation detectors and associated electronic data acquisition systems is experimental testing and qualification in a well-characterized and representative radiation field in a reference irradiation facility. The neutron flux levels in modern material testing reactors (MTRs) are in the range of 1E14 – 1E15 n cm-1 s-1. However, the number of dedicated test facilities in Europe is currently decreasing, with research reactors recently and soon-to-be shut down.
The 250 kW JSI TRIGA reactor is a very well characterized reactor in terms of the knowledge of the neutron and gamma fields, a product of the work performed at the JSI over the last decade, mostly in collaboration with the Instrumentation, Sensors and Dosimetry Laboratory at CEA, Cadarache. Therefore it fulfils very well the first criterion as a reference facility. However, in steady state operation it is able to generate a maximum neutron flux level of around 2E13 n cm-1 s-1, i.e. several orders of magnitude lower than the MTR-relevant range. In steady-state mode the requirement of representativeness is therefore not fulfilled well. On the other hand, the JSI TRIGA reactor can operate in pulse mode, due to its prompt negative temperature coefficient of reactivity. In pulse operation, one of the control rods of the reactor – the transient rod – is ejected to a pre-set height, thus introducing a sufficient amount of reactivity to make the reactor prompt supercritical. The reactor power increases very quickly to a peak value, up to around 1 GW, after that the reactor power drops, due to the negative reactivity which is a consequence of the elevated fuel temperature. Depending on the inserted reactivity, the pulse duration is of the order of a few seconds (low and long pulses), to 5-10 milliseconds (high and short pulses). The neutron flux level is proportional to the reactor power level, therefore the highest attainable flux is nearly 1E17 n cm-1 s-1, albeit for a short amount of time.
In 2019, a bilateral collaboration project between the CEA and JSI was initiated, to investigate the possibility of neutron flux measurements performed at very high neutron flux levels in reactor pulse operation, made possible by a modern, validated, wide dynamic range neutron acquisition system. The project aims at demonstrating the feasibility of nuclear instrumentation and associated electronic data acquisition system tests at the JSI TRIGA reactor at neutron flux levels relevant to MTRs.
This paper presents the first measurements in reactor pulse operation, performed during an experimental testing campaign in collaboration with researchers from the CEA, and in collaboration with the Instrumentation Technologies (I-Tech) company, using the I-Tech-developed current meter. An experimental campaign is scheduled to be carried out at the JSI TRIGA reactor jointly by CEA and JSI researchers in the autumn of 2020 using the CEA-developed MONACO fission chamber data acquisition system.

08.09.2020 15:40 Poster Session

Research reactors - 306

Reactivity initiated accident (RIA) analysis of ITU TRIGA Mark II research reactor using PARET/ANL

Ozge Ozkan1, Senem Senturk Lule2, Uner Colak2

1Istanbul Technical University, Energy Institute, 34469 Maslak, Istanbul, Turkey

2Istanbul Technical University, Energy Institute, Ayazaga Campus, 34469, Istanbul, Turkey


ITU TRIGA Mark II is an open pool, water moderated research reactor and has a steady state power of 250 kW. It is located in the Energy Institute of Istanbul Technical University, Turkey. In this study, Reactivity Initiated Accident (RIA) analysis of ITU TRIGA Mark II research reactor is carried out for different reactivity insertion cases using PARET/ANL code. PARET/ANL is a computer code that couples the thermal hyraulics and point kinetics equations. In addition, it is used for transient analysis of research reactors. Since it is necessary to observe the behaviour of the reactor in case of any transient, thermal hydraulic safety analysis plays an important role for nuclear reactor safety. Analyses of $0.09 slow ramp reactivity insertion in 1 second and, $1.35 and $1.5 fast ramp reactivity insertions in 0.5 seconds are performed for ITU TRIGA Mark II research reactor in the scope of this study. These reactivity insertions are analysed applying scram condition while the reactor is under steady state operation. Scram point is set to %110 of nominal power which is 275 kW as indicated in safety analysis report. Power, total reactivity, fuel centreline, clad surface and coolant temperature variations with time, and minimum DNBR (Departure from Nucleate Boiling Ratio) values are observed using PARET/ANL after reactivity insertions. The simulation results showed that the upper limits of power, fuel centreline, and clad surface temperatures are not exceeded and minimum DNBR values are within the safety limits. The research reactor operates safely in case of aforementioned ramp reactivity insertions according to results of this study.

08.09.2020 15:40 Poster Session

New reactor designs and small modular reactors - 401

TEPLATOR: Basic design of the primary circuit

Michal Zeman1, Anna Fortova2, Radek Skoda1

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


The TEPLATOR is an original way of district and industrial heating using nuclear power. It means, that TEPLATOR uses spent nuclear fuel from nuclear power plants. The spent nuclear fuel is the one that did not reach its regulatory and design limits. This fuel can be taken either from spent fuel pool or interim storage. This means that the fuel for TEPLATOR is already manufactured and thus no additional cost for fuel arises.

This one of a kind design will be demonstrated in a demonstration unit “TEPLATOR DEMO”. This DEMO unit has 50 MW of thermal power with 55 spent fuel assemblies of VVER-440 in the core. The fluid output temperature from the core is 98 °C, thus the whole unit works on atmospheric pressure. It is constructed as three loop system with three main pumps and three heat exchangers. This paper describes the basic design of primary circuit. The main idea concerning the design is explained first. Then the evolution of the design from the first steps to current 3D model is included. Next discussion is focused on individual components of the TEPLATOR (i.e. heat exchangers, pumps etc.) Briefly the construction and operation of compensation means are presented. Finally, the whole concept of TEPLATOR DEMO and its design is summarized.

08.09.2020 15:40 Poster Session

New reactor designs and small modular reactors - 402

Basic design of the TEPLATOR core - construction

Jiří Závorka1, Radek Skoda2, Martin Lovecky3

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

2Czech Technical University, Zikova 1903/4, 166 36 Prague 6, Czech Republic

3University of West Bohemia, Univerzitni 8, 306 14 Plzeň, Czech Republic


The study shows the base optimization of the TEPLATOR core. One of the most difficult challenges for this innovative concept is dealing with irradiated fuel assemblies. Because spent nuclear fuel has insufficient reactivity, the main aim of this study is to investigate various effects on TEPLATOR operation from the perspective of the core design. The analysis was executed by Serpent code and shows the influence of individual components of the TEPLATOR CORE. The crucial role plays the choice of suitable moderator; it determines the construction fundamentals of the active zone. Based on this choice an ideal fuel pitch, a dimension of a reflector and parameters of cooling were arranged. The construction with or without fuel channels was dealt with. After consideration of all these effects the first core of this kind was designed.

The first DEMO is designed with 50 MW of thermal power and 55 spent fuel assemblies of VVER-440 type in the core, heavy water as both moderator and coolant. More is described in the article.

08.09.2020 15:40 Poster Session

New reactor designs and small modular reactors - 403

TEPLATOR: Basic economic study for the construction and operation

David Masata1, Jana Jirickova2, Radek Skoda2

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


Competitiveness of district and process heat production based on fossil fuels is challenging. Costs of fuel, technology improvement for reaching the emission limits and carbon credits purchasing increase price of produced heat. The TEPLATOR concept is an innovative way to eliminate all these cost by using the spent nuclear fuel for future zero emission district heating.
This article is focused on economics of the TEPLATOR concept. The objective of study was to investigate economic feasibility of the project. Initially, the summary of worldwide heat production and consumption and overview of district heating systems is provided for a study of TEPLATOR application. Then complex construction costs study has been carried out and evaluated. Financial analysis for TEPLATOR operation is presented considering fueling, maintenance, district heating systems requirements and safety parameters. All the financial aspects have been compared with conventional district heating plants. Finally, the feasibility was summarized and total investment costs and produced heat price were determined.

08.09.2020 15:40 Poster Session

New reactor designs and small modular reactors - 405

Possible implementation of ex-core measurement in TEPLATOR graphite reflector

Eva Vilimova, Tomas Peltan, Jana Jiřičková

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


Ex-core neutron flux measurement is crucial system for all common power reactors. It is necessary to monitor neutron flux and control the chain reaction therefore ex-core neutron flux measurement is one of the main safety and control system. The main advantage of this arrangement of detectors is fast response on neutron flux change, which determines the reactor power change. For all new SMR concepts it is crucial to improve detection systems suitable for these reactors. Most of the SMR reactor concepts are based on graphite moderator or reflector, which is also the case for TEPLATOR. The TEPLATOR is very innovative solution of district heating system based on heavy water as a moderator and graphite as a reflector. TEPLATOR is designed to use irradiated fuel from commercial PWR, VVER or BWR reactors, which has low to intermediate burnup. This article focuses on verification of possible use of special measuring system placed in graphite reflector. For calculations performed in this article was used Monte Carlo code Serpent.

08.09.2020 15:40 Poster Session

New reactor designs and small modular reactors - 406

Natural uranium as alternative fuel for TEPLATOR

Tomas Peltan, Eva Vilimova, Radek Škoda

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


The TEPLATOR is an innovative solution for district heating using nuclear energy. It is a specially designed critical assembly with a specific arrangement of the core with 55 fuel assemblies which are moderated and cooled by heavy water and operated at atmospheric pressure with low output temperatures compared to commercial nuclear power plants. The TEPLATOR DEMO is designed for using irradiated fuel from PWR, VVER and BWR reactors. In cases that the irradiated fuel is not available in terms of high burnup or other reasons, there is a possibility to fuel it with natural uranium. This article focuses on development of natural uranium core within the TEPLATOR design. It is mainly concerned with neutronic development of fuel assemblies with suitable parameters for this application. This article contains various fuel modifications with different time of operation. All calculations were performed by Monte Carlo code Serpent.

08.09.2020 15:40 Poster Session

New reactor designs and small modular reactors - 407

ENEA Preliminary Results for the Participation to the NEA/EGIF Fuel Codes Benchmark Phase II

Rolando Calabrese

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


The NEA Expert Group on Innovative Fuels (EGIF) has started a follow up of the benchmark on fuel performance codes that has been concluded recently. Phase II will focus on the thermomechanical response of sodium fast reactor fuel pins. The main interest of this exercise is on predictions during unprotected transients. Codes’ results will be used to identify the level of agreement of the models applied in the codes.
The outcomes of Phase II are expected to benefit the initiatives of the NEA Expert Group on Uncertainty Analysis in Modelling (EGUAM). This group has extended the results achieved by the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force. The analyses conducted by EGUAM have been mostly devoted to calculate global neutronic quantities, reactivity feedback coefficients, and unprotected transients simulations. A deeper evaluation of uncertainties affecting these quantities will be discussed in the near future and, in this view, the information coming from the EGIF benchmark should be helpful for a more sound understanding of uncertainties.
This paper presents some preliminary calculations performed by ENEA for the Phase II of the benchmark. Moving from the determination of the reactivity feedback coefficients by EGUAM, the thermomechanical behavior during an unprotected transient of the fuel pin of a large size oxide core is studied by means of the TRANSURANUS fuel performance code. This large size core (3600 MWth) is cooled by sodium and is composed of 453 fuel subassemblies and two enrichment zones. The driver fuel is MOX with an Oxide Strengthened Steel (ODS) cladding. Geometrical and materials specifications together with modelling approaches are discussed in view of a more detailed definition of a Phenomena Identification and Ranking Table (PIRT).

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 501

Use of ADVANTG to analyse skyshine dose rates around DEMO

Domen Kotnik, Bor Kos, Igor Lengar

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


DEMOnstrational power plant, in short DEMO, is currently being developed within the EUROfusion Power Plant Physics and Technology Department [1]. It is assumed to be constructed around 2050 and will represent the next milestone for the realization of commercial fusion power plants, whereby the main goal will be a demonstration of electricity production. To ensure that the regulatory radiation dose limit, due to neutrons and gammas, is not exceed at DEMO site boundary at any moment, adequate shielding is required. In this work, we systematically analysed different DEMO scenarios, from normal operation to maintenance scenarios after shutdown, and optimized the shielding performance provided by the external wall and the domed roof of the maintenance hall. Furthermore, the skyshine effect, i.e. scattering from the surrounding air back to the ground, was also considered. In addition, the impact of different parameters on the skyshine effect, e.g. air humidity, ground density and surrounding landscape types, were examined as it may influence the required thickness of tokamak building walls depending on the choice of geographical location. Due to extreme complicity of both deep penetration/shielding problem and large radiation skyshine effect, all the calculations were performed using two-step hybrid transport method utilizing MCNP [2] for particle transport and ADVANTG [3] for variance reduction.


1. Federici G. et al., 2019 Nucl. Fusion 59, 066013
2. C.J. Werner, et al., "MCNP6.2 Release Notes", Los Alamos National Laboratory, report LA-UR-18-20808 (2018).
3. S.W. Mosher et al., ADVANTG – An Automated Variance Reduction Parameter Generator, ORNL/TM-2013/416 Rev. 1, Oak Ridge National Laboratory, 2015.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 502

Method to investigate induce eddy currents in spherical tokamaks

Shahab-Ud-Din Khan

national tokamak fusion program , islamabad, 3329, Pakistan


In this paper, theoretical approach has been applied to investigate induce eddy currents in central solenoid and vacuum vessel. Central Solenoid (CS) is a key component that providing the inductive voltage to initiate and sustain the plasma current. It has also provided the position and shape of the plasma. The peak and area under the curve of Current profile is depended on the charging voltages, self-inductance, mutual inductance, resistance of the coil, and the eddy current produced in the central pipe and vacuum chamber. Therefore, to deal eddy current issue we have studied some important numerical methods. Furthermore, we have developed numerical methods based on coupling schemes and then find out the required solution for induced eddy current in central solenoid and vacuum vessel.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 503

TCV tokamak neutron shielding assessment and upgrade

Bor Kos1, Henri Weisen2, Patrick Blanchard2, Jerémie Dubray2, Basil Duval2, Duccio Testa2, Matteo Vallar2, Luka Snoj1

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

2Swiss Federal Institute of Technology (EPFL), Station 3, Lausanne, Switzerland


The Tokamak a configuration variable or Variable configuration tokamak (TCV) [1,2] is a magnetic fusion research device located at the Swiss Plasma Center, EPFL. Its unusual rectangular vacuum vessel, 1.5m high and 0.5m wide equipped with tightly fitted poloidal field coils allows for a wide variety of plasma shapes to be investigated for properties such as energy confinement and magnetohydrodynamic stability.
Plasma electron heating up to ~3 MW is provided by a set of gyrotrons operating in the range 80-120 GHz, while heating of the ions up to Ti~3 keV is currently achieved with a 25-30 keV, 1.4 MW neutral beam heating (NBH). A second NBH system operating at 50-60 keV is planned to enter operation early 2021. When operating in deuterium (D), the NBH with a single 25 keV produces 2.45 MeV D-D neutrons at rates of several 1012 n/s, depending on plasma conditions. Radiation exposure in the nearby control room currently meets strict safety limits (1 ?Sv/day). However, when operated together in D with the 2nd beam, also in D, neutron rates are predicted to increase by an order of magnitude, exceeding safety limits (unless hydrogen is used, which is disadvantageous for ion heating experiments).
The insufficiency of the current radiation shielding for optimal deuterium operation of both NBH units has motivated a design study for an upgraded shielding of the experimental hall. Neutral particle (neutron and gamma) transport simulation were performed in order to identify the main particle streaming pathways and propose additional shielding to reduce the neutron and gamma dose to an acceptable level. To perform the analysis the state-of-the-art hybrid (deterministic/stochastic) particle transport methodology was adopted. The hybrid methodology combines the ADVANTG [3] code for determining efficient variance reduction parameters based on a rough deterministic transport simulation with a high fidelity continuous energy stochastic particle transport simulation performed in the second step using MCNP [4].
The shielding analysis was performed in several steps. In the initial steps the most effective material for shielding was determined and the effect of gaps between shielding blocks was quantified. After that, the existing TCV hall MCNP model was upgraded with major components of the TCV including the toroidal and poloidal coils, vacuum vessel, support pillars and wooden floors. After performing simulations with the present shielding configuration a basic shielding design was proposed and analyzed using the hybrid transport methodology. The MCNP computational design will serve as the basis for upgrades to the current shielding.
[1] Hofmann et al "Creation and control of variably shaped plasmas in TCV." 1994.
[2] Coda, S., M. Agostini, R. Albanese, S. Alberti, E. Alessi, S. Allan, J. Allcock et al. "Physics research on the TCV tokamak facility: from conventional to alternative scenarios and beyond." Nuclear Fusion 59, no. 11 (2019): 112023.
[3] Mosher, Scott W., Aaron M. Bevill, Seth R. Johnson, Ahmad M. Ibrahim, Charles R. Daily, Thomas M. Evans, John C. Wagner, Jeffrey O. Johnson, and Robert E. Grove. "ADVANTG—an automated variance reduction parameter generator." ORNL/TM-2013/416, Oak Ridge National Laboratory (2013).
[4] Kiedrowski, Brian, Thomas E. Booth, Forrest B. Brown, Jeffrey S. Bull, Jeffrey A. Favorite, R. Arthur Forster, Roger L. Martz, and M. C. N. P. Documentation. "MCNP5-1.6, Feature Enhancements and Manual Clarifications." LA-UR-l0-06217 (2010).

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 504

Tritium retention in and permeation through the FW of the DEMO reactor

Olga Ogorodnikova

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


A fundamental understanding of hydrogen (H) isotope migration and retention in neutron-irradiated tungsten (W) and advanced steels in the presence of a temperature gradient, stress field and dynamic surface modification is crucial for assessment of tritium retention and permeation, as well as for developing methods for removing tritium. In addition, tritium control is necessary for material performance and the plasma fuel balance. The rate-equation model is successfully used for simulation and interpretation of wide range of available experimental data of deuterium depth profile measurements, thermal desorption spectra, erosion in the case of deuterium (pure and with seeding) interaction with tungsten and different kinds of steels from gas state, atomic state and plasma. For predictions, it is necessary to know (i) parameters of particle-material interaction such as diffusion, binding energies of H isotope with different kinds of defects, recombination, reflection and sputtering coefficients, etc. and (ii) input parameters such as ion energy and ion flux of each kind of incident particles, heat flux, etc. in stationary regime and transients. Here, the physical parameters and model assumptions are validated by comparison with wide range of laboratory experiments and AUG and JET data. A comparison of laboratory data with tokamak data using same materials and same methods of measurement gives us a platform for understanding of key parameters govern H isotope retention in present tokamaks and reliable extrapolations towards future fusion devices. Experimental and simulated deuterium (D) concentrations at radiation-induced defects as well as total D retention are in very good agreement. The model's predictions of 1999-2000 years are confirmed by subsequent experimental data of 2009-2017 years, indicating deep understanding of the underlying physics and the reliability of the physical parameters included in the model. Impurity effect also included in the present model by dynamic surface composition change during plasma exposure. It is shown that the He seeding into the D plasma reduces the D retention, while the Be (also C, N or O) seeding increases the D retention that is in agreement with experimental data. The model is validated also against experimental data of ELMs simulations of W and Eurofer. It is shown that the synergetic effect of D, He and high heat flux leads to completely different particle retention and material modification compared to separate/sequential irradiation. Calculations show that the tritium (T) retention in the bulk of all metallic first wall increases with increasing the ELM frequency and power, leading to the trapping of T in radiation defects in the colder region. Direct extrapolation to DEMO from available experimental data is not possible, since the synergetic effect of incident ions, neutron irradiation and high heat flux leads to completely different particle retention, permeation and material modification compared to irradiation in available facilities. The extrapolation requires modelling.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 505

Effect of D on the creation of displacement damage in tungsten during multiple sequential irradiations with MeV W ions

Matic Pečovnik1, Sabina Markelj1, Thomas Schwarz-Selinger2

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

2Max-Planck-Institut für Plasmaphysik (IPP), Boltzmannstr. 2, D-85748 Garching b. München, Germany


In future tokamak fusion reactors, displacement damage of the material lattice caused by 14 MeV fusion neutrons will significantly increase D retention compared to a pristine material. The creation and evolution of displacement damage can be affected by many factors, such as temperature and irradiation fluence.

In this study the effect of D on displacement damage creation during W irradiation was studied. For this purpose, three bulk tungsten samples were used. The first sample was irradiated with 20 MeV W ions at room temperature up to a calculated damage dose equal to 0.23 dpa. Afterwards, the created displacement damage was decorated using a low energy D plasma at 370 K sample temperature. For the second W sample this W irradiation and D decoration sequence was performed twice with the same irradiation/exposure conditions. The experimental results of the first two samples have shown a substantial increase in the amount of created displacement damage of nearly a factor of two [1]. To further elucidate the mechanism behind this increase, a third sample was sequentially irradiated three times. The D concentration depth profiles were measured using Nuclear Reaction Analysis and the D desorption spectra were determined using Thermal Desorption Spectroscopy. A further increase in the amount of displacement damage is observed.

The analysis of the third sequential W irradiation/D exposure allowed not only the study how D presence affects creation of displacement damage, but also how different D concentrations present in the material during irradiation affect displacement damage creation. To further elucidate the before mentioned effects, a displacement damage creation and stabilization model [2] included in the macroscopic rate equation code MHIMS-R [3] was used to recreate the experimentally measured D depth profiles and desorption spectra.

Despite using only a few sequential W irradiations and D exposures, the use of our model allowed us to extrapolate our experimental results to a large number of sequential W irradiation/D exposure steps. Furthermore, this allowed us to determine the amount of displacement damage created in a quasi-stationary regime involving many sequential irradiation/exposure steps which is equivalent to a simultaneous irradiation and hence will more closely resemble conditions prevailing in a tokamak. The model predicts, that in such a quasi-stationary regime, the trapped D concentration increases from 1.7 at.% when no D is present to 4.2 at.% when D is present during W irradiation.

[1] T. Schwarz-Selinger, J. Bauer, S. Elgeti, and S. Markelj, “Influence of the presence of deuterium on displacement damage in tungsten,” Nucl. Mater. Energy, vol. 17, pp. 228–234, Dec. 2018.
[2] M. Pečovnik, E. A. Hodille, T. Schwarz-Selinger, C. Grisolia, and S. Markelj, “New rate equation model to describe the stabilization of displacement damage by hydrogen atoms during ion irradiation in tungsten,” Nucl. Fusion, vol. 60, p. 036024, 2020.
[3] E. A. Hodille et al., “Study of hydrogen isotopes behavior in tungsten by a multi trapping macroscopic rate equation model,” in Physica Scripta, 2016, vol. 2016, no. T167, p. 014011.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 506

IMASViz: a general data visualization software for IMAS

Dejan Penko

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


IMASViz is a software utility for visualization and data analysis of the static and dynamic data, stored within the data structures of the Integrated Modelling Analysis Suite (IMAS). IMAS is a scientific software framework infrastructure that coordinates the collective development and execution of integrated plasma codes and plasma applications describing fusion operations in experiment devices such as ITER, JET, and WEST tokamak.
The suite foundations are set on an underlying Physics Data Model (PDM) which facilitates the basis for coupling the plasma physics codes with IMAS standardized database structures named Interface Data Structures (IDSs).

IMASViz provides convenient access to the archived data through Graphical User Interface (GUI) which reflects the Physics Data Model (PDM) and utilizes Univeral Access Layer (UAL) for accessing the data to IDS pulse files written in MDSPlus format or UDA for remote access. It allows easier and more straightforward browsing through the contents of the IDSs which resemble a tree data hierarchy. No prior knowledge of the PDM is required.
The application allows straightforward data vizualization support for 0D and 1D signals. Some 2D data visualization support is provided through the use of plugins, while other 2D and 3D visualization utilities are presently under development. IMASViz allows the display of data for different cases at once based on Multiple Data Interface (MDI), this way enabling efficient and straightforward data comparison between different scenarios. Furthermore, IMASViz provides an easy plugin integration for adding new features or more specific utilities.

IMASViz application is written in Python3.7 programing language and utilizes Python libraries such as PyQt5 for GUI, Matplotlib and PyQtGraph for visualization functionalities, and Sphinx for generating HTML and PDF documentation that include a complete user manual.
The project is being hosted at the central repository maintained at the ITER Organization (IO).
Presently, IMASViz is being deployed on GateWay HPC, ITER HPC, and WEST HPC where it is extensively used by EUROfusion consortium members. While the application itself is already available for use it is still under active development, and various features, GUI improvements, etc. are still being implemented. In this article, IMASViz, its major features and its significance to the ITER Integrated Modelling programme will be presented.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 507

Time-dependent Boundary Conditions in ITER Scrape-Off-Layer

Ivona Vasileska

University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva 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.

The effects caused by the wall erosion represent the boundary conditions in regions of plasma-surface interaction and the limiting expressions for the parallel heat flux and viscosity. The formulation of boundary conditions (BCs) and their time dependence is an interesting and important task for plasma edge studies.

The aim of this work was to derive time-dependent BCs at the plasma wall transition (PWT) for ELM-free, Type-I ELM and post-ELM states based on a kinetic test simulation in the ITER tokamak with neutrals. As this is a time-consuming process, the simulations are first conducted without neutrals, that was done at the previous article in NENE 2019, and then the neutrals are added to the system.

This contribution describes the first results of attempts to address this issue for ITER simulations under high-performance conditions using the 1D3V electrostatic parallel Particle-in-Cell (PIC) code BIT1. 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 the inner to the outer target, assuming typical upstream separatrix parameters of n~3-5x10^19 m^-3, T_e~100-150 eV and T_i~200-300 eV. Inclined magnetic fields at targets of (~5) are included, as are the particle collisions, with a total of 3.4 x 10^5 poloidal grid cells, giving shortening factors of 20. At a later stage, these will be used as BCs for the calculations of the ELM target heat loads using the SOLPS-ITER code.

A typical simulation requires up to 60 days running massively parallel 1152-2304 cores of the EU Marconi super-computer. The duration of the ELM pulse is taken to be between 100-400 mikros.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 508

Preliminary simulation of actively cooled divertor mono-block made of W-based composite

Patrik Tarfila1, Boštjan Končar1, Matej Tekavčič1, Saša Novak2, Petra Jenuš3, Matej Kocen3

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

2Jožef Stefan Institute, Department for nanostructured materials, Jožef Stefan International Postgraduate School, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

3Jožef Stefan Institute, Department of Nanostructured Materials K-7, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Plasma-facing components (PFC) of the divertor will be exposed to severe conditions (extremely high heat fluxes, thermal shocks, erosion, etc), which will affect the structural integrity of the materials and may cause serious damage of the component during operation. Target plate in DEMO (Demonstration fusion power plant) divertor must be capable to withstand intense power loads between 10 and 15 MW/m2 during steady-state operation. To remove such high heat flux densities, the component needs to be actively cooled using an efficient cooling design and it has to be made of the state-of-the-art materials for the PFC and heat sink structures. Development of durable and efficiently cooled divertor target requires common research efforts in the field of materials and in the design of cooling solutions.

Recently, a new W-composite material has been developed at Jožef Stefan Institute, within the EUROfusion WPMAT project. The new material, developed specifically for the use in DEMO divertor target plate, has a high melting temperature (similar to that in the pure tungsten) and is additionally toughened by nano-size W2C particles. The material samples have already been characterized in terms of its thermal and mechanical properties. The machining and preparation of actively cooled mock-up using the JSI material is currently underway at Karlsruhe Institute of Technology (KIT). The mock-up includes the mono-block made of W-composite and a CuCrZr cooling pipe with a swirl tape as the turbulence promoter. The target plate will be assembled from many of such mono-blocks. The proposed mono-block cooling solution and the new PFC material must go through a number of tests, numerical and experimental, to demonstrate its suitability for DEMO conditions.

To evaluate the effect of heat loads on the temperature distribution and pressure drop in actively cooled mono-block, the preliminary numerical simulations have been performed using the ANSYS CFX code prior to the actual experimental campaign. The boundary conditions of the model are selected to match the capabilities of the High Heat Flux (HHF) experimental facility JUDITH-2 in FZ Jülich. In the present study the computational model based on experimental mock-up is presented. A coupled fluid-solid approach is used to solve the heat transfer between the solid parts and the coolant flow. The turbulent flow in the coolant is modeled by a Shear Stress Turbulence (SST) model using a steady-state simulation. The simulation results are presented, compared with empirical correlations and adequately discussed.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 509

From research to semi-industrial production of tungsten-based monoblocks

Matej Kocen1, Petra Jenuš1, Anže Abram1, Eliseo Visca2, Saša Novak3

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

3Jožef Stefan Institute, Department for nanostructured materials, Jožef Stefan International Postgraduate School, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


Future fusion reactor DEMOnstration Power Station will demand better materials, especially inside the reactor in the proximity of plasma. The known materials will not withstand the harsh environment over longer period of time, due to higher temperatures and greater loads that are present in reactors nowadays. That is why tungsten-based composites have been greatly researched in the last decades.
Throughout research, we have prepared W-based composites with small ditungsten carbide (W2C) particle inclusions [1, 2]. The material exhibited great mechanical properties at room temperature as well as at elevated ones. Mechanical properties do not deteriorate greatly even if the material is exposed to 2000 °C for 24h in a vacuum. Specific samples were also tested under high-heat fluxes. The microstructure of the specimen revealed, that only small cracks were formed when the material was exposed to the highest loads. Comparing to other reference materials (i.e. “ITER-grade tungsten”), our material performed excellently. This gave us green light for the next step; fabrication of real-size monoblock, which will be tested under similar conditions as in the DEMO reactor.
With upscaling the process new obstacles appeared. So far, all samples were cylindrically shaped due to the limitations of the consolidation technique, thick just a few millimeters and with a diameter of a maximum of 25 mm. But monoblock is in the shape of a cuboid with dimensions 23x28x12 mm3. Cylindrical samples with a diameter of 40 mm and thickness of more than 12 mm had to be made and additionally machined. Hard and tough materials such as tungsten and its composites cannot be machined by conventional methods. The only way is to use Electrical discharge machining (EDM). The brass wire in the machine cuts through and disintegrate the material using high-frequency electrical spark discharges. Finishing surface was covered in melted oxides, that is why it had to be again ground and polished. The inner hole of monoblock will be attached to CuZrCr cooling pipe, therefore crucial element of this part was wetting of the composite surface with the pipe. Wetting tests of differently treated surfaces were tested in ENEA Frascati Research Center.

Key words: tungsten, ditungsten carbide, DEMO divertor, monoblock, Electrical discharge machining, wetting

This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. Parts of the work have been performed within the PhD studies of Mr Matej Kocen supported within the EUROfusion education & training scheme. This project has received funding from the Slovenian Research Agency (Contracts No. 1000-17-0106, J2-8165, P2-0087-2 and P2-0405-5).

1. Jenus, P., et al., W2C-reinforced tungsten prepared using different precursors. Ceramics International, 2019. 45(6): p. 7995-7999.
2. Novak, S., et al., Beneficial effects of a WC addition in FAST-densified tungsten. Materials Science and Engineering: A, 2020. 772: p. 138666.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 510

Formation of an Inverted Sheath in front of an Eledtron Emitting Electrode that Terminates a Bounded Plasma System Studied by a PIC Simulation using the XPDP1 Code

Tomaz Gyergyek1, Jernej Kovačič1, Stefan Costea2

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


Potential formation in front of electrodes that emit electrons has been a hot topic of investigations in plasma physics at least from first applications of emissive probes. More recently this topic is becoming more and more important also in fusion research since it is becoming clear that first wall in ITER made of tungsten will very probably be strongly electron emitting at least in the divertor region. Relatively recently it has been predicted theoretically that at a very strong electron emission from a floating electrode, the floating potential of such an electrode could become positive with respect to the plasma potential and a so-called inverted sheath is the formed in front of the electrode. Recently a one-dimensional kinetic model of an inverted sheath in a bounded plasma system has been developed. In this work the same problem is studied further with particle-in-cell simulations. The code XPDP1 is used for this purpose. From the source electrons and singly charged positive deuterium ions are injected, while from the collector only electrons are injected. All 3 groups of particles are injected with half-Maxwellian distributions with different temperatures and injection fluxes. It is shown that stable potential profiles which decrease monotonically from the collector to the source can be obtained. The distribution functions have cut-off shapes, as assumed in the model. It is clear that the simulations are in good qualitative agreement with model. A more detailed quantitative comparison is planned for the future.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 511

Assessment of power deposition on plasma facing components inside WEST tokamak with the use of field line tracing

Matic Brank1, Mehdi Firdaouss2, Marie-Helene Aumeunier2, Gregor Simič3, Leon Kos1

1University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta 6, 1000 Ljubljana, Slovenia

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

3Faculty of Mechanical Engineering, University of Ljubljana, Aškerčeva 6, 1000 Ljubljana, Slovenia


The WEST tokamak is a French tokamak that originally began to operate as Tore Supra. The original name came from the words torus and superconductor, as Tore Supra was for many years the only tokamak of this size with superconducting toroidal magnets. After a major upgrade to install tungsten walls and divertor, the tokamak was renamed WEST (W-Environment in Steady-state Tokamak). The main aim of the upgrade was to create a ITER-like walls in order to have a test facility for ITER divertor and wall components. Magnetic coils have been added as well to confine the originally circular plasma into and ITER-like "D"-shape. As one of the main testing tokamaks for ITER, it is thus important to provide assessment of heat loads on plasma facing components inside WEST.

This article describes a series of benchmarks performed between two magnetic field line tracing codes, PFCFlux and SMITER. The code PFCFlux (Plasma Facing Components Flux) has been developed for heat flux calculations on those components, including shadowing effects. Its main purpose is the evaluation of the power deposition on the poloidal limiters (high and low field sides) and the divertor targets. SMITER is a graphical user interface (GUI) framework, built around SMARDDA kernel for magnetic fieldline tracing, for power deposition mapping on tokamak plasma-facing components (PFC) in the full 3-D CAD geometry of the machine.

The input parameters to field line tracing codes are the target geometry, shadowing geometry and 2-D equilibrium data. Target geometry describes the geometry where the power fluxes will be assessed. Field lines are traced back from the target triangles into the scrape-off layer region. Shadowing geometry acts as a shadow. If field lines of a target triangle hit the shadowing geometry, then this triangle isconsidered non-wetted. If the field line does not hit anything after a user-defined maximum distance, then the target riangle is considered wetted, and power deposition can be calculated based on the provided heat flux profile. Equilibrium data contains information about magnetic field fluxes, that are usually defined in $R-Z$ space and are assumed to be axisymmetric in the toroidal direction. Another optional input parameter is the 3-D magnetic field data, that can define the ripple in the magnetic field, which affect the wetted area and power deposition calculation. The magnetic field line tracing studies, presented in this article, deal with both axisymmetric cases, as well as presence of the ripple in the magnetic field.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 512

Transport calculations in circular symmetric geometries – application to fusion tokamak sector models

Igor Lengar, Andrej Žohar, Domen Kotnik

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


When neutron transport calculations are performed in circular symmetric geometries, the numerical models most frequently cover only one sector of the whole geometry and reflecting boundary conditions are used on both edges, on which the sector is truncated in order to mimic the full geometry. This is the case also for the majority of calculations for the largest fusion reactors – the large tokamaks. The approach is applicable if the radiation source is also circular symmetric. In the case of localized sources the sector models in unaltered way do not give correct results anymore since the symmetry among the sectors brakes down and the flux in each one has to be treated individually.

A possible way to solve the problem when using localized sources is to expand the numerical model in order to fill up the full 360°. This can, however, be very difficult for complex models. In the present work another approach was followed in which the model is left unaltered and the correctness of results is assured by a modification of the transport code. In this way correct results can be obtained in a much quicker way by using only one sector of geometry and instead modifying the transport code by additionally labelling the particles upon reflection of the edges of the modelled sector. The detailed concept and the needed modifications to the transport code are described on the example of the Monte Carlo code MCNP and on a simplified 22.5° sector model of a tokamak geometry. The approach can be used for quick calculations for the case when only sector models are available, as in case of the large tokamaks. It is shown that the standard way of using an average result over all sectors gives for localized sources an inferior result to the described method.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 513

Calorimeter design for Neutral Beam Injector of DTT – Thermo-hydraulic analysis

Domen Ovtar1, Boštjan Končar2, Oriol Costa Garrido3, Martin Draksler2, Piero Agostinetti4

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

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

4Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Izpolni naslov!, Padova, Italy


NENE Abstract
Calorimeter design for Neutral Beam Injector of DTT – Thermo-hydraulic analysis
*1,2D. Ovtar, 1O. Costa, 1M. Draksler, 3P. Agostinetti, 1B. Končar
1Jožef Stefan Institute, Ljubljana, Slovenia
2Faculty of Mechanical Engineering, University of Ljubljana, Slovenia
3Consorzio RFX, Padua, Italy
The main steps in the development of the calorimeter design for the Neutral Beam Injector (NBI) of Divertor Test Tokamak (DTT) are presented. One of the design drivers has been also the cost reduction, hence the presented concept uses CuCrZr panels as the heat absorbing components, similar to the ones used in the Residual Ion Dump (RID) of the ITER Neutral Beam Test Facility (NBTF). The beam power absorbed in the panels is removed by the pressurized water flowing through the cooling pipes drilled into the panel structure. The heat transfer to the coolant is enhanced by twisted tape inserts. The design and dimensions of the panels and cooling pipes are based on the thermal-hydraulic optimisation using detailed Computational Fluid Dynamics (CFD) analyses.
The analysis considers local thermal-hydraulic parameters, such as maximum temperature of the panels and of the coolant, coolant pressure drop and geometric constraints like panel inclination angle and available space for example. The power of the DTT NBI beam takes into account the beam imprints and power distributions in the beam core and in the beam halo, which are modelled by superposition of Gaussian functions. In CFD simulations the coupled solid-fluid heat transfer problem is solved. Main design choices and optimisation are determined by minimising the heat transfer coefficient at the interface between the panel structure and the highly turbulent coolant flow.

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 514

Self-consistent model of thermionic emission from the divertor tiles

Stefan Costea1, Jernej Kovačič2, Tomaz Gyergyek2

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


The scrape-off-layer plasma in the tokamak acts as the main exhaust for impurities as well as residual heat from the core plasma. All modern tokamaks now have the scrape-off-layer (SOL) plasma connected to the solid surfaces through a configuration named the divertor. Divertor spreads the plasma heat flux over a larger area and enables a much better impurity pump-out.
Nonetheless, the divertor material will be pushed to the limits in future tokamaks and its allowed temperature will be one of the main limiting factors of the operational scenarios [1]. The heating and cooling of the divertor leads to material damage, melting, tritium retention etc. But in recent years the predictions show that parts of divertor (or even whole) could become so hot, that thermionic emission might start playing a role. Thermionic emission of electrons is a process in which the surface starts emitting a significant number of electrons, if the temperature is high enough. This process is governed by the Richardson-Dushman law and is highly non-linear, but until recently not much attention has been put to it in the fusion community [2]. Now we believe, that it might be a very important mechanism, that could either lead to improved cooling or to additional heating of the divertor surface [3].
In order to better understand what role the thermionic emission could play in the behaviour of the SOL plasma, we have coupled a fully-kinetic particle-in-cell code BIT1 [4] to a simplified thermal model of a divertor monoblock. The kinetic description of the SOL plasma is the most accurate possible and has the best predictive capabilities for the heat flux calculations. In this way we were able to produce a self-consistent model of electron emission based on the balance of surface cooling sinks and heating sources. The source was parametrically modelled according to the expected SOL power flows during pre-, inter- and post-ELM. The first results show the significant influence that the thermionic emission has on the formation of the plasma-wall-transition layer and consequently on the heat loads to the divertor.

[1] J. P. Gunn et al., Nucl. Fusion 57 (2017), 046025
[2] M. Komm et al., Plasma Phs. Control. Fusion 59 (2017), 094002
[3] M. D. Campanell, Phys. Plasmas 27 (2020), 042511
[4] D. Tskhakaya et al., J. Nucl. Mat. 438 (2013), S522-S525

08.09.2020 15:40 Poster Session

Nuclear fusion and plasma technology - 515

20 years of Slovenian neutronics activities within EU fusion programme

Ivo Aleksander Kodeli

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


This year marks the 20s anniversary since Slovenia joined the European Fusion Program and 15 years since the establishment of the Slovenian Fusion Association. In June 2000 the first Slovenian fusion project was obtained within the European Commission entitled "Benchmark experiments to validate European Fusion data". The activities in the area of fusion neutronics performed in these 20 years in the scope of the different EU fusion organisations, EFDA, Fusion for Energy and EUROfusion, will be summerised. Among others, our contribution to the benchmark experiments performed and/or ongoing with the EU partners at the FNG facility at ENEA Frascati will be presented.

* present address UKAEA, UK

08.09.2020 15:40 Poster Session

Radiation and environmental protection - 601

A New Paradigm on Plastic Waste

Andrej Trkov

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


Plastics of various kind constitute an important material for a wide variety of products that we use in everyday life. In fact, plastics are such versatile materials that it is hard to imagine any substitute.

Practically every product at the end of its useful lifetime becomes waste. Plastic waste is considered a severe environmental problem because it is insoluble and practically undegradable. Some of it is recycled, but a significant fraction of it is disposed of either by incineration, or simply dumped into the environment. Incineration itself is potentially problematic, because some types of plastic may produce toxic fumes.

We can look at plastic from a completely different perspective. Practically all plastics are organic compounds that are based on carbon. Thus, plastics can be viewed as a way of carbon storage. We can still use plastics in products for everyday life, but at the end of their useful lifetime we treat them as carbon sequestration. Most of the plastics use natural oil as the raw material. Returning carbon from the oil into disposal facilities would at least make plastics carbon-neutral. In the long run, plastics can also be made by hydrogenation of carbon dioxide extracted from the air and polymerization.

The key point in this proposal is to have cheap electricity from non-fossil sources for pumping power to extract CO2 and for the electrolysis to produce hydrogen. The inevitably higher cost of plastics produce in this way would limit their use for non-essential purposes, but it would also require government action to restrict plastic production from fossil raw materials. The required energy demand cannot be met solely from existing technologies of renewable due to their low power density, which would require unacceptably large land areas devoted to this purpose. From this point of view, nuclear energy seems the obvious natural approach to the negative carbon footprint in the use of plastics.

08.09.2020 15:40 Poster Session

Radiation and environmental protection - 602

Distribution Methods of Potassium Iodide Tablets to be Used as a Protective Measure in Case of a Major Nuclear Reactor Emergency

Metka Tomažič, Anja Grabner, Saša Kuhar

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


Iodide prophylaxis is one of the urgent protective measures recommended by several international organizations and adopted by several countries worldwide in case of nuclear accidents. Prophylactic use of potassium iodide ensures that the thyroid gland is fully loaded with non-radioactive iodide and therefore blocks the uptake of radioactive iodine thus reducing internal radiation exposure of the thyroid gland, which does not discriminate between radioiodine and non-radioactive iodide. This is possible if the potassium iodide is taken before or shortly after an exposure to radioactive iodine and if taken in the proper dose.
Since the intake of potassium iodide needs to be timely, the distribution of potassium iodide tablets needs to be planned ahead. Even though countries in Europe and worldwide have adopted the World Health Organization recommendations of potassium iodide doses for different risk groups homogeneously in their emergency response plans - also because of the implementation of the new EU Basic Safety and Standards Directive - BSSD, 2013/59/Euratom - the implementation of iodide prophylaxis in terms of distribution are still quite specific to each country.
Methods of distribution in European countries vary and in practice two methods are used, pre-distribution and stockpiling. In each country both methods face certain country-specific difficulties in implementing them. In example, in Slovenia, the pre-distribution of potassium iodide tablets in the urgent protection zone has not been very successful in the past few years.
The paper therefore discusses the logistics of pre-distribution of potassium iodide tablets in Slovenia and other European countries and searches for best practices by analysing and interpreting primary (legislature) and secondary sources (scientific and journalistic works, data from different databases and results of questionnaires), with the objective to develop recommendations for improved pre-distribution methods in Slovenia.

Key words: iodide prophylaxis, intake of potassium iodide, protective measures, nuclear and radiological accidents, distribution, pre-distribution.

08.09.2020 15:40 Poster Session

Radiation and environmental protection - 603

Serpent 2 validation for radiation shielding applications

Silja Häkkinen


Serpent 2 is a multi-purpose three-dimensional continuous-energy Monte Carlo particle code, developed at VTT [1]. Serpent’s capabilities include among others reactor physics applications, multi-physics simulations and neutron and photon transport simulations. Serpent is actively developed and the validation of Serpent’s various capabilities is a continuous effort. The purpose of this work is to contribute to the validation of Serpent’s photon transport and coupled neutron-photon transport routines and variance reduction techniques.

Photon transport mode was first introduced in Serpent version 2.1.24 in 2015 [2]. The present version 2.1.31 includes four main photon interactions and three secondary photon production mechanisms [3]. The main interactions are photoelectric effect, Rayleigh scattering, Compton scattering and electron-positron pair production. The secondary mechanisms are electron-positron annihilation, atomic relaxation and bremsstrahlung emitted by electrons and positrons. Photonuclear reactions have been implemented in a development version 2.1.32 [4]. A weight-window based variance reduction scheme has also been implemented in the current distribution of Serpent [5].

In this work, the afore mentioned methods are validated with appropriate benchmark calculations. The ICSBEP benchmarks ALARM-CF-FE-SHIELD-001 and ALAMR-CF-PB-SHIELD-001 involving neutron and photon spectra calculations of a Cf-252 source through different sized iron and lead spheres are calculated. Other benchmarks such as e.g. SINBAD benchmarks may also be considered. Serpent calculations are compared to MCNP6.2 calculations and measurement results.

1. Leppänen, J., et al. (2015) "The Serpent Monte Carlo code: Status, development and applications in 2013." Ann. Nucl. Energy, 82 (2015) 142-150.
2. Kaltiaisenaho, T., “Implementing a Photon Physics Model in Serpent 2”, (Master’s thesis), Aalto University, 2016.
3. Kaltiaisenaho, T. (2015) “Photon transport physics in Serpent 2 Monte Carlo code”, Computer Physics Communications, In Press, (2020).
4. Kaltiaisenaho, T. (2019) “Photonuclear reactions in serpent 2 Monte Carlo code”, M&C 2019, Portland, USA, 25.-29.8.2019.
5. Leppänen, J., Viitanen, T., Hyvönen, O. (2017) ”Development of a Variance Reduction Scheme in the Serpent 2 Monte Carlo Code”, M&C 2017, Jeju, Korea, 16.-20.4.2017.

08.09.2020 15:40 Poster Session

Radiation and environmental protection - 604

Coupling of ASTEC and RASCAL Codes to Evaluate the Source Term and the Radiological Consequences of a Loss-of-Cooling Accident at a Spent Fuel Pool

Antonio Guglielmelli1, Stefano Ederli2, Pietro Maccari2, Federico Rocchi1, Fulvio Mascari1

1Italian National Agency for New Technology, Energy and Substainable Economic Development, Via Martiri di Monte Sole, 4 - Bologna , 40129, Italy


The Laboratory for the Safety of Nuclear Installation of the Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) is involved in the Management and Uncertainties in Severe Accidents (MUSA) H2020 European Project, coordinated by CIEMAT. Within the MUSA project, ENEA is involved in the Innovative Management of SFP Accidents Work Package (WP6), coordinated by IRSN. In this WP ENEA is committed to perform a Severe Accident analysis on a Fukushima-like Spent Fuel Pool (SFP) by means of the deterministic ASTEC V2 code, owned by IRSN, with the aim to apply innovative measures on the SFP Severe Accident Management to mitigate the Radiological Consequences (RC) of the accident itself. This paper, developed by ENEA as a preparatory study for the MUSA Project WP6, presents the RC results of a Loss-of-Cooling SA at a Fukushima-like SFP coupling the ASTEC V2 and RASCAL 4.3 codes. Specifically, the Source Term (ST) provided by ASTEC is used as input for the RC analyses. The SFP model used for the ASTEC analysis is an upgrade of the one adopted in the AIR-SFP European Project, and it will be further developed by ENEA to be used within the MUSA project. The RC were evaluated through the atmospheric dispersion of the ASTEC ST with the Gaussian puff dispersion module of the RASCAL 4.3 code, owned by U.S. NRC. In order to perform the RC studies, the Fukushima-like SFP is assumed located in one of the Italian cross-border NPP sites. The weather data adopted are hourly meteorological data on the chosen site for the most conservative day, with respect to the RC on the Italian territory, in a ten years range (2002-2011). The results of the RC for 96 hours of ST release from the SFP in a range of 160 km from the emission point are reported in terms of Total Effective Dose Equivalent (TEDE), I-131 thyroid dose and Cs-137 total ground concentration. The mitigating effect on ST and RC of the cooling spray systems actuated with several pH values was also investigated.

08.09.2020 15:40 Poster Session

Radiation and environmental protection - 605

Analysis of the HI-TRAC Transfer Cask Dose Rates During Nuclear Power Plant Krsko Spent Fuel Dry Storage Campaign One

Paulina Dučkić, Davor Grgić, Mario Matijević, Radomir Ječmenica

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


In this paper shielding analysis is performed to establish neutron and gamma dose rates around the transfer cask HI-TRAC loaded with Fuel Assemblies (FA) from Nuclear Power Plant (NPP) Krsko Spent Fuel Dry Storage (SFDS) Campaign one. The HI-TRAC is a multi-layered cylindrical vessel designed to accept a Multi Purpose Canister (MPC) during loading, unloading and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis is divided into two steps. The first step is the source term generation using ORIGEN-S SCALE sequence. The source is calculated based on the operating history of spent fuel assemblies currently located in NPP Krsko spent fuel pool. The obtained intensities and spectra of the spent fuel assemblies are used in the second step to calculate the dose rates around the transfer cask. A comprehensive shielding analysis included the calculation of dose rates resulting from fuel neutrons, fuel gammas, neutron induced gammas (n-g reaction), and hardware gammas under normal and accidental conditions. To obtain the dose rates with the acceptable uncertainties, FW-CADIS based variance reduction is adopted, as implemented in ADVANTG code. The dose rates around HI-TRAC cask are calculated using MCNP6 code for all 16 casks loading belonging to Campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization (Holtec International).

08.09.2020 15:40 Poster Session

Radiation and environmental protection - 606

Neutronic design of the first built experimental cave for the European spallation source

Tamás Bozsó1, Eszter Dian2, Szabina Török3

1Budapest University of Technology and Economics, Műegyetem rkp 3-9, Budapest 1111, Hungary

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


The Neutron Macromolecular Diffraction (NMX) Experimental Cave is being built as first of a kind in the European Spallation Source (ESS) in Lund, Sweden. The experimental cave walls provides the biological shielding to personnel standing outside its walls. This implies modelling for solid walls and estimate the effect of chicanes and cable inlets. Concerning the layout, the basic shape is an ‘L’, which is designed to contain a labyrinth to protect the entrance door. The door is located in front of the labyrinth. The door can be accessed from a separate landing structure equipped with lift for pallets and stairs. There are two separate concrete beams to support the robot rails which are designed to bear the loads of the robots moving the detectors inside the cave.
The cave and its door have to comply with radiological constraint of 3 µSv/h contact dose rate at every 20cmx20cm surface area. For the simulation a disk neutron source was defined in the neutron guide entrance, providing a neutron beam with 2 cm diameter, hitting the sample or any assumed target. Simulations were performed with a normalised incident neutron spectrum 3.4×109 neutron/s beam intensity. The detailed study using MCNP6 simulation of NMX experimental cave shielding confirmed that the proposed version with 90 cm concrete wall thickness and a 65 mm steel door behind a blade wall provides sufficient shielding at all the operational (H1) and accidental (H2) scenarios. The simulated dose rates outside of the experimental cave do not exceed the predefined 1.5 µSv/h (safety factor 2) limit for the Monte Carlo simulations and 1 µSv/h (safety factor 3) limit for the analytical calculations.

08.09.2020 15:40 Poster Session

Radiation and environmental protection - 607

Assessment of safety barriers and effectiveness of instant control of the subcriticality level of fissile materials’ clusters localized inside the NSC-SO

Serhii Kupriianchuk1, Roman Godun2, Kostiantyn Sushchenko2

1Institute for Safety Problems of NPP’s, National Academy of Sciences of Ukraine, 36a Kirova str., Chornobyl, Kyiv reg.,Ukraine, 07270,, Ukraine

2POWER, Izpolni naslov!, USA


According to the NSC-SO’s (New Safe Confinement “ARCH” and “Shelter” object) current status, its nuclear safety (NS) is provided by safety barriers (SB), as well as through the organizing and technical measures. After NSC installation into its design position (November 2016), humidity and temperature storage conditions of unorganized fissile materials’ (FM) inside the NSC-SO were changed. Also, after November 2016 a significant increase of neutron activity was recorded at the periphery of clusters of potentially nuclear-dangerous FM (CPNDFM). In view of the foregoing, there was an urgent need for the SB re-evaluation and verification. Within the comprehensive experimental and computational studies the declared SB were evaluated and it was shown that they are either absent or require the verification under the NSC conditions. Thus, the statement about the SB presence in the NSC-SO is not true.
The effectiveness of the instant control of CPNDFM' subcriticality level (which implements by introducing of the solutes of neutron-absorbing materials (NAMs) by NSC-SO control systems) was also evaluated. It was found that after the operation of these systems the NAMs' concentration on accessible periphery of CPNDFM is not enough for the effective control of subcriticality level of this cluster. In addition, lava-like fuel-containing materials (localized on this cluster's periphery) act as a filter of NAMs, which prevents their entering into the CPNDFM epicenter.
Evidence that using of state NS control systems of NSC-SO is ineffective (for NS purpose) is presented. This statement is mainly based on the fact that these systems supply the NAMs' solutes only to the CPNDFM periphery, since there are no direct access' routes directly to the epicenter of this cluster. Also, according to results of the studies of water flows routes in localization zone of the CPNDFM, it was found that on their way exists a cascade of natural pools (formed during the accident). The cascade of natural pools on the routes of NAMs solutes, firstly, determines the delay of neutron absorber's entering into the CPNDFM localization zone (since this water pool's cascade needs to be gradually filled), and secondly, it determines the minimum volume of NAMs' solution that must be estimated experimentally.
Recommendations on the improving of instant control's effectiveness of CPNDFM' subcriticality level are provided.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 701

Scale effects’ determination methodology for buoyant impinging jets

Benjamin Jourdy1, David Guenadou2, Michel Gradeck3, Alexandre Labergue3, Nathalie Marie4

1CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France

2CEA Cadarache, DTN/STRI/LHC, Bar. 728, FR-13108, Saint Paul lez Durance, France

3LEMTA Université de Lorraine CNRS , 2, av. de la Foret de Haye - TSA 60604, 54518 Vandoeuvre les Nancy, France

4CEA Cadarache, IPSN/DRS/SEMAR BAT. 700, 13108 St. Paul Lez Durance, France


The CEA is involved in the development of 4th generation sodium cooled reactors for which specific codes were developed. However, they need to be validated using experimental results obtained on relevant mock-ups. Due to the complexity of building a full-sized prototype in the nuclear field, most of the experiments are performed on reduced-sized models but it may lead to scale effects. We have to ensure that the validation carried out at small scale is effective at the reactor one. As a critical issue in SFR reactor is the rising of the jet outgoing the core at low power, the scale effect on this phenomenon is studied.
This article is a detailed review and methodology presentation of the scale effects and thermo-hydraulics study of this complex phenomena based on the Richardson’s similarity. The dimensionless Reynolds-Averaged Navier-Stokes equations (RANS) under Boussinesq’s approximation and the Vaschy-Buckingham theorem are applied to determine the relevant dimensionless numbers.
Then, a scale effect study methodology based on the scale series is detailed. The already existing mock-up MICAS is used as the reference scale, and a new mock-up called MOJIT-EAU has been designed to be representative of the smaller scales. This mock-up can be modified to allow scale effects investigation from 1/6 to 1/3 of MICAS’s scale, but also allow a phenomena investigation to quantify the influence of the complex geometry on the raise of the jets.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 702

Large-Eddy Simulation of the TRIGA Mark II Reactor Core

Carolina Introini1, Antonio Cammi2, Andrea Salvini3

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

3Laboratorio Energia Nucleare Applicata Universita degli Studi di Pavia, Via Aselli 41, 27100 - Pavia, Italy


Complex industrial applications still handle turbulence modelling using Reynolds-Averaged Navier Stokes (RANS) models. This approach reduces the required computational effort by eliminating all turbulent fluctuation through the time-averaging process. Therefore, fluid equations are in terms of local-averaged flow field quantities. With RANS, all turbulence is modeled; the main issue is that there is no universal RANS model for turbulent behaviour, nor there is a fundamental physical law underlying all RANS models. A certain model may be adequate for a specific application but unsuitable for another. Despite these limitations, RANS is still the most popular approach in use for nuclear safety analysis.  

Large-Eddy Simulation (LES) approaches lie between Direct Numerical Simulation (DNS) and RANS. Whereas DNS directly reproduces all the scales of turbulence without any modelling assumption, the LES approach resolves only the larger scales of turbulence, while still modelling the smaller ones. This method uses a low pass filtering based on grid size on the equations, eliminating the smaller length and time scales. Compared to DNS, this approach reduces the required computational effort.

Turbulence modelling remains a key point of Computational Fluid-Dynamics analysis. Accurate simulations must take into account the impact that turbulence can have on the diffusion of momentum and energy. For wall-bounded flows such as the cooling channels of nuclear reactors, additional turbulence terms may cause a significant increase in heat transfer and wall stresses. Turbulent phenomena may also be responsible for instabilities. An accurate safety analysis of nuclear systems requires a very precise and rigorous treatment of turbulence. Even though RANS models still represent the state-of-the-art in this field, the feasibility to more accurate approaches, in light of the fast improvements in terms of computational speed, represents an ongoing field of research. This is especially true for more complex reactor designs where turbulence may play a significant role.

Reactor modelling using LES is still a field of research. This work further explores this field by proposing a full LES analysis of the reactor core of the TRIGA Mark II reactor. The limited dimensions of this system make it ideal for evaluating how full LES computations perform for a whole reactor core, both in terms of accuracy and needed computational resources. This work presents some preliminary results of the evaluation, also highlighting the mesh adjustments needed to optimize the simulation. 

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 703

Requirements of LES in natural circulation loops: preliminary study in a pipe geometry

Angelo Battistini1, Antonio Cammi2, Marco Colombo2, Micheal Fairweather3, Carolina Introini4, Stefano Lorenzi5

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

2Politecnico di Milano - department of energy, Via La Masa 34, 20156 Milano, Italy

3LEEDS, Izpolni naslov!, United Kingdom

4Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

5Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy


In available computational fluid dynamics (CFD) literature, large eddy simulation (LES) has been mostly employed in reproducing channel flow behaviours. Instead, less attention has been given to the analysis of flows in axial-symmetrical geometries such as a pipe flow, mainly because of the common knowledge that LES requirements and predictive capabilities in such systems could be derived from channel flow results.
In this work, LES is applied to the simulation analysis of a pipe flow, in a preliminary step to evaluate the applicability of the method to study the stability of natural circulation loops. Therefore, pipe diameter and Reynolds number are chosen to be representative of those encountered in the experimental facility DYNASTY, built and operated at Politecnico di Milano. The facility has been specifically designed to study natural circulation with fluids with internal heat generation, and the reliability of passive safety strategies in the Generation IV molten salt reactor. The Reynolds number in particular is found to be rather low in DYNASTY, and often on the boundary of the laminar to turbulent transition. This, even more under natural circulation, is known to be particularly challenging for Reynolds-averaged Navier-Stokes turbulence models, traditionally employed in the CFD study of similar systems. Therefore, here simulations are carried out with a LES approach, using different meshes of increasing quality and refinement. The flow is incompressible and adiabatic at a Reynolds number Re=5300, in a pipe with cyclic inlet and outlet boundary conditions and with diameter D = 0.038 m. The length to diameter ratio is L / D ~ 13, which has been found to be high enough to allow the propagation without excessive damping of the turbulence structures. In all the simulations, the WALE eddy-viscosity subgrid-scale (SGS) model has been used to model the unresolved part of the turbulence spectrum. This choice was driven by its documented adaptability to a large range of Reynolds numbers, specifically important under natural circulation, and its improved performances with respect to the standard Smagorinsky SGS model. The spatially- and temporally-averaged results of the simulations are compared against reference literature data from direct numerical simulation. In addition, given their importance for the stability features of natural circulation loops, assessment of the prediction of pressure drops against the Darcy-Weisbach equation is also made.
The simulation results show a good capability of the model in predicting turbulent structures, alongside a positive trend towards accuracy when the quality of the discretization grid is increased. As a major outcome, an important dependence of the main turbulent variables (e.g. friction losses, averaged turbulent shear stresses) from the discretization parameters in the radial, tangential and longitudinal directions is found, pointing out the importance of such analysis. Minimum refinement levels necessary to achieve a certain accuracy are identified, this being critical to ensure the model will remain computationally sustainable when applied to the large spatial scales and long temporal transients required in natural circulation loop studies.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 704

Statistical Uncertainty of Turbulent Heat Flux in a Flow of low-Prandtl fluid over a Backward Facing Step

Jure Oder, Iztok Tiselj

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


In this paper, we present the profiles, together with statistical uncertainties of turbulent heat flux, obtained in the recently performed Direct Numerical Simulation (DNS) of flow of two liquid metals over a confined backward facing step. This work is a continuation of work that was performed within the European project SESAME. The presented profiles are obtained in the whole domain, while the uncertainty is analysed in 49 pre chosen points throughout the domain. The analysis is performed over at most 10 million time step or equivalently of around 5000 dimensionless time units. The turbulent heat flux is important for development of new models with less accurate methods for simulating flow properties, such as methods, where the Navier-Stokes equations are averaged.
The DNS was performed with the Nek5000 code. The most notable feature of this code is the use of spectral elements to solve for velocity, temperature and any other passive scalar. It is an open source code developed by the Argonne National Laboratory.
Spectral element method is a hybrid method between a finite element method and a collocation spectral method. The method divides the computational domain into finite elements, within which a spectral method is used to solve for variables. This method allows for the use of spectral method in irregularly shaped geometries and to perform direct numerical simulations in such geometries.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 705

Prediction of the flow and thermal field in a complete PWR fuel assembly using a reduced-resolution RANS

Blaž Mikuž1, Ferry Roelofs2

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


Reproduction of a single-phase turbulent flow and heat transfer inside a pressurized water reactor (PWR) fuel assembly is a challenging task due to the complex geometry and the huge computational domain. In order to tackle this problem, the capability of a reduced-resolution CFD approach is examined together with the application of polyhedral computational cells, which are known for their ability to fit well around a complex geometry and at the same time require fewer cells for the same accuracy. The accuracy of the considered numerical method had been first assessed against measurements in a 5×5 rod bundle with mixing grid. Then, the method has been expanded to a larger computational domain consisting of an array of 15×15 fuel rods and a single mixing grid. The obtained predictions revealed the interesting pattern of swirl flow as well as diagonal cross flow downstream the mixing grid, which is driven by the applied design of split-type mixing vanes. In the present paper, the computational model is extended to a domain representative of a complete PWR fuel assembly with ten mixing grids, inlet and outlet sections. Pressure drop and flow field in the given fuel assembly are analyzed together with the predicted temperature field and potential hot spots. In spite of a relatively coarse spatial resolution of the applied approach, the reduced-resolution CFD provided promising results at least for the qualitative prediction of the pressure, flow field and location of hot spots.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 706

Numerical simulation and validation of vortex shedding frequency in a vortex flowmeter.

Jan Sotošek1, Boštjan Končar2

1Institute of Physics Faculty of Natural Sciences and Mathematics, Gazi Baba bb, 1000 Skopje, Macedonia

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


Numerical simulation and validation of vortex shedding frequency in a vortex flowmeter

Jan Sotošek*1,2, Boštjan Končar1
1Jožef Stefan Institute
2Faculty of mathematics and physics, University of Ljubljana

*Corresponding author:

Vortex flowmeters are widely used in various industries, including nuclear, due to their excellent characteristics, such as high accuracy, linear output signal, wide measurement range, absence of moving parts and low cost of investment and maintenance. The essential part of the vortex flowmeter is an obstacle (bluff body) that is mounted transversely in the measuring pipe. Behind a bluff body, vortices are generated periodically, with a shedding frequency that is directly proportional to the volume flowrate. However, a clear linear dependence usually appears only at higher pipe Reynolds numbers above 20000.

In this study, the flow behind the bluff body is simulated numerically over a range of Re numbers, between 5000 and 50000. The simulations are compared and validated against the dedicated experiments performed in an air pipe with a specially designed prismatic bluff body. Detailed numerical simulations should provide a deeper insight into the vortex shedding phenomena and help us to understand the disruption of linearity at lower Reynolds numbers.

Simulations are carried out using the open-source code OpenFOAM. In order to model turbulent vortex shedding flow over a wide range of volume flowrates, different simulation techniques need to be used. At lower Re numbers, Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) methods can be used. DNS is able to solve the the Navier-Stokes equations directly, but requires a very fine numerical mesh. LES can be solved on a slightly coarser mesh, but needs additional turbulence model (WALE in our case) to resolve turbulence dissipation on the sub-grid scale. Both methods are computationally too demanding at higher Re numbers above 30000, where the Unsteady Reynolds Averaged Navier Stokes (URANS) simulations with with k-? SST turbulence model is used.

The simulated vortex shedding frequencies are validated against the available experimental data, where the frequency signals behind the bluff body were obtained using a hot-film probe. Several different techniques have been tested and compared, to obtain unique frequency characteristics from the simulations. Pressure, velocity, lift force and vorticity oscillations at three different points behind the bluff body have been recorded, fast Fourier transform or zero crossing algorithm have been used to determine the frequency.

Likewise, a comprehensive verification of simulations has been performed to ensure sufficient accuracy and convergence of numerical results. Different numerical methods require different numerical schemes that have been tested, showing that DNS needs most accurate higher-order schemes. Cyclic boundary condition ensures fully developed velocity profile at the pipe inlet.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 707

Heat transfer measurements and visualization of subcooled flow boiling of R245fa in a narrow horizontal annular duct

Boštjan Zajec, Boštjan Končar, Leon Cizelj

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


Subcooled flow boiling was investigated in a temperature-controlled horizontal annular test section. The installed experiment is a part of the laboratory THELMA (Thermal Hydraulics experimental Laboratory for Multiphase Applications) built at Reactor Engineering Division of Jožef Stefan Institute. The unique characteristics of the temperature-controlled heating of the inner tube in principle enable a wide range of measurements, from subcooled boiling to the critical heat flux and above into the unstable film boiling regime. Internal thermocouples allow measurements of local surface temperatures and heat fluxes, while the transparent outer tube allows the visualization of flow boiling phenomena.

In this paper, the influence of bubble parameters on boiling heat transfer is studied. Subcooled flow boiling of R245fa is recorded with a high-speed camera at constant inlet temperature and varied mass flow rate of the heating-water. Videos of the flow boiling patterns are analysed frame-by-frame, using the image processing, to determine bubble size distributions. Measurements of local thermocouples in the inner tube have been used to obtain the heat transfer characteristics. The experimental setup, analysis methods and preliminary results are presented.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 708

Experimental and numerical analysis of heat transfer on the annular test section in single-phase flow regime

Boštjan Zajec, Anil Kumar Basavaraj, Blaž Mikuž, Boštjan Končar, Marko Matkovič

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


A water-heated annular test section for studies of convective boiling was recently installed in the Thermal-Hydraulics Experimental Laboratory for Multiphase Applications (THELMA). The test section is designed to investigate various fluid flow and heat transfer regimes in reactor-like conditions. As the total heat flux is calculated from the temperature difference between inlet and outlet thermocouples, accurate determination of heat losses on the test section is required. Our previous experimental studies in single-phase have shown that surface heat losses are non-negligible and are difficult to estimate from the inlet and outlet temperatures.

For better understanding of the phenomena involved, a detailed three-dimensional numeric model of the test section and its inlet/outlet heads was designed in ANSYS Fluent. Two limiting cases with small and large temperature difference between water and working liquid (refrigerant R245fa) were simulated. Experimentally measured axial temperature profile and outlet temperatures were used to validate the simulation results, while numerically obtained velocity and temperature fields are used to determine the areas of largest heat losses. The results and a qualitative model of heat losses are presented in this paper.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 709

Application of artificial neural network for prediction of two-phase flow parameters

Feride Kutbay, Senem Senturk Lule, Uner Colak

Istanbul Technical University, Energy Institute, Ayazaga Campus, 34469, Istanbul, Turkey


The two-phase columns are widely used by air conditioner systems, heat exchanger and evaporation systems, and petroleum refinery systems. The advantages of bubble column which rely on two phase phenomena are good heat transfer without leakage and being environmentally friendly and cost effective due to small amount of wear and tear. Therefore, investigations on two-phase flow is a crucial aspect to achieve high energy efficiency. The analysis of two-phase flow parameters such as phase fractions, temperature, velocity can be carried out numerically and experimentally. Nonetheless, the complex nature of gas is a difficult to analyze with computational fluid dynamics (CFD) based on calculation tools without having high computing power within a short period of time. There is a tendency to use artificial neural networks (ANN) in fluid dynamics and heat transfer research to achieve fast and accurate solution without extended computing requirements.
In this study, neural network and CFD solver hybrid methods were coupled. The heat transfer coefficients and pressure drop measurements for six different mass flux, two different saturation temperatures, and the specified range of heat flux from vertical column filled with R134a experiment were used as testing data sets. The experimental set up was modelled by CFD code Fluent to acquire the training data sets. The network then trained with 1296 simulation results as a design parameter. Once trained, the network was benchmarked against the test data. Benchmark results are in good agreement. Therefore, ANN can be utilized for the prediction of parameters difficult to measure in such systems.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 710

Large Eddy Simulation of turbulent flow with different Prandtl numbers near uniformly heated wall

Jan Kren, Blaž Mikuž, Iztok Tiselj

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


At Reactor Engineering Division of Jožef Stefan Institute a new experiment is being built in the THELMA laboratory to study temperature fluctuations at the wall cooled by adjacent turbulent flow. As the experiments are limited especially in resolution and also in the available insight into the flow and thermal field, Computational Fluid Dynamics (CFD) studies are widely used as a design support to fluid dynamics experiments.
A new experimental section is being built that consists of a 3x3cm2 and roughly 4m long square channel, with a 59cm long part of the wall heated with a thin foil. The flow and heat transfer in the new experiment have been reproduced with a wall-resolved Large Eddy Simulation (LES) using a wall-adapting local eddy-viscosity (WALE) model in OpenFOAM computer code. Simulations assumed incompressible, fully developed turbulent flow of Newtonian fluids with Reynolds number of 10000 and two different Prandtl numbers corresponding to water and liquid lead. Heat transfer has been included to the study using temperature as a passive scalar, so there is no feedback loop on mass and momentum conservation equation. This approximation is sufficient if the bouyancy effects are not large. However, results for water coolant have shown that the temperature increase next to the heated plate is more than 50 degrees for the given power of 1kW at the heated wall. Thus, the study suggests that a reduced power will have to be applied in the experiment in order to avoid too high temperature in the system.
Moreover, we are particularly interested in distribution of temperature in turbulent regime. For example, with observation and analysis of coherent thermal structures in simulations we have predicted the required resolution and frequency needed by thermographic high-speed camera at given Reynolds and Prandtl numbers.

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 711

Investigation of Temperature Fluctuation in Conjugate Heat Transfer

Mohit Pramod Sharma, Blaž Mikuž, Iztok Tiselj

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


Conjugate heat transfer both in single and two phase fluid flow are widely encountered in
day to day life and have industrial importance too. In conjugate heat transfer, temperature
fluctuation of fluid adjacent to solid wall plays important role in system design and safety.
The temperature fluctuations at fluid-wall interface depends on thermal activity ratio; which
is the ratio of product of thermal capacity and thermal conductivity of fluid to solid.
However, there are several computational databases on excessively simplified geometries,
which have not been verified with the experiments till date. Also, only handful experimental
studies on temperature fluctuation in thin solid wall exist in literature.
In the view of this, a new experimental facility has been designed and installed at Reactor
Engineering Division of Jožef Stefan Institute. The aim of present study is to investigate
thermal fluctuation on heated foil in case of square duct. In this paper, design of experimental
facility and diagnostics used such as thermocouples, IR camera, flow meter etc. has been
presented in details.
Keywords: Temperature fluctuations, Conjugate heat transfer, Thermal activity ratio

08.09.2020 15:40 Poster Session

Thermal-hydraulics, computational fluid dynamics - 712

Assessment of CHF models for single side heating of ITER-type divertor

Mohit Pramod Sharma1, Vinay Menon2, Samir Khirwadkar2

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


PFC (Plasma Facing component) is an important component of the divertor, a region in the
tokamak whose main function is to serve as a plasma exhaust and is subjected to very high heat
loads. Hence it is important to cool divertor for effective operation and to prevent any damage to
the PFCs. This is accomplished with pressurised water operating in the subcooled region
extracting heat out of divertor. The importance of subcooled water is that it increases the heat
capacity and overall efficiency of the system. However, the Critical Heat Flux (CHF) must be
accurately predicted while designing PFCs for safe operation.
The major focus of this work is to evaluate sub-cooled CHF models against the latest
experimental CHF data reported for various divertor geometries under single sided heating conditions.
Keywords: CHF, Single Sided heating condition, PFC, ITER Divertor

08.09.2020 15:40 Poster Session

Safety analyses - 801

Thermo-mechanical analysis of a dry storage system

Julio Benaviddes Rodriguez1, Oriol Costa Garrido2, Gonzalo Jimenez1, Leon Cizelj3

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

2Institut "Jožef Stefan", Jamova cesta 39, 1000 LJUBLJANA, Slovenia

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


As the spent fuel pools in nuclear power plants are near their full capacity and the construction of centralized sites, such as Yucca Mountain (USA) or Villar de Canas (Spain), remains unclear, the nuclear industry is moving towards in-house storage of spent fuel using dry casks. Together with the extension of the lifespan of nuclear reactors, this has increased the need to understand the behavior of nuclear spent fuel during its storage phase in a dry cask system to ensure that safety limits are not compromised.
One of the safety parameters is the peak cladding temperature (PCT), which should not surpass 400°C as per the NRC ISG-11 rev.3 - Interim Staff Guidance. In order to calculate this temperature, simulations using computational fluid dynamics (CFD) codes are employed, as these have proven to reliably predict temperature distributions inside dry storage systems. However, one of the biggest uncertainties when calculating PCT in a dry cask system is the gap size between the basket and the steel canister. The gap size uncertainty originates from manufacturing and building tolerances, and from the difficulty of calculating thermal expansion on a CFD simulation. Previous works have modelled the gap as a thermal resistance. These have found that, depending on the gap size in the dry cask system TN-24P, PCT values may increase up to 28°C.
In this paper, the temperature distribution of the TN-24P dry cask obtained in a CFD simulation is imported into the finite element (FE) ABAQUS code to perform the thermo-mechanical analysis of the cask. The TN-24P cask is chosen due to the experimental data available in the open literature. The goal of the paper is to obtain the thermal expansion and mechanical stresses of the casks components. The results of the simulation show that the gap size is not uniform along the cask’s height and, overall, rather low mechanical stresses. This paper thus shows that a better understanding of dry cask systems can be obtained from the successful coupling of CFD and FE codes.

08.09.2020 15:40 Poster Session

Safety analyses - 802

Updating of Quench - 12 experiment calculations using ASTEC2.2b computer code

Antoaneta Stefanova, Pavlin Petkov Groudev

Institute for Nuclear Research and Nuclear Energy, 72 Tzarigradsko chaussee, Blvd., BG-1784 Sofia, Bulgaria


The present work presents an update of Quench 12 experiment calculations using ASTEC V2.2b computer code. The main purpose of the QUENCH program is to investigate fuel behavior in severe conditions. The main object of the QUENCH program is to examine the behavior of overheated fuel under different flooding conditions and to create database for model development and improvement of Severe Fuel Damage computer code packages.
The QUENCH-12 experiment has been performed to investigate the behavior of VVER fuel assemblies with a hexagonal lattice and fuel rods with claddings made by Zr1%Nb (E 110), which is used in VVER reactors. The ZrO2 pellets as in the LWR-type tests represent the fuel.
The test was conducted at Forschungszentrum Karlsruhe on 27 December, 2006 in the frame of EC-supported ISTC program.
The recently issued (Accident Source Term Evaluation Code) ASTEC2.2b computer code has been use for performing of an update of Quench 12 test calculations. The study is focuse on investigation of quenching of overheated VVER fuel behavior, bundle oxidation processes and hydrogen generation. The calculated results has been analysed and compared with QUENCH12 test data.
The simulation have been perform in the frame of EC, ASCOM project.

08.09.2020 15:40 Poster Session

Safety analyses - 803

Stress Analysis in Nuclear Reactor Pressure Vessel of VVER 1200 Nuclear Power Reactor

M A Rashid Sarkar


Sadia khan, Sadman Sakib, Subrata saha and M A RASHID SARKAR
MIST DHAKA Bangladesh

A reactor pressure vessel (RPV) contains the fuel assembly control rods coolant, core shroud etc. The reactor vessel body is designed to contain the fuel assembly, coolant, and fittings to support coolant flow and support structures. It is usually cylindrical in shape with both ends ellipsoidal in geometry. The RPV provides a critical role in safety of the PWR reactor and the materials used must be able to contain the reactor core at elevated temperatures and pressures. The materials used in the cylindrical shell of the vessels have evolved over time, but in general they consist of low-alloy ferritic steels clad with 3-10mm of austenitic stainless steel.
In this paper the simultaneous effects of mechanical,thermal and gamma radiation-induced stresses in RPV of VVER 1200 NPP are considered. Mechanical stress arises due to internal pressure of the RPV. In addition, thermal stress and gamma radiation-induced stress are also considered to simulate the actual condition. The thermal stresses are developed in a solid body whenever the expansion or contraction of a differential volume element, which would normally result from a change in temperature, is prevented. Thermal stress in RPV must be considered with also the internal heat generation in the wall due to absorption of gamma radiation from fission. The absorption of gamma-ray in the wall of RPV is modeled here as exponential internal heat generation in long thick-walled cylinders.
Triaxial Stress State is used for determining the principal stresses of the Finite Elements of RPV wall material. ANSYS Finite Element Method is used to simulate the stress state. Exactly the same diameter and material of the shell, operating temperatures, internal pressures and gamma ray absorption rate of the RPV of VVER1200 are considered here for stress analysis and compared with the data available in the literature.

The paper is for presentation in NENE20 conference

08.09.2020 15:40 Poster Session

Safety analyses - 804

Accuracy quantification with FFTBM-SM of BETHSY 9.1b test simulations by TRACE and RELAP5

Andrej Prošek, Boštjan Končar

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


The TRAC/RELAP Advanced Computational Engine (TRACE) is today the state-of-the-art and one of the world’s leading best estimate systems codes in the field of thermal-hydraulics. It is intended for safety analyses of loss-of-coolant accidents and operational transients, and other accident scenarios in pressurized light-water reactors (PWR) and boiling light-water reactors. For TRACE code assessment, the 9.1b test performed on Boucle d'Études Thermo-Hydraulique Systeme (BETHSY) has been selected, representing beyond design basis accident with non-degraded core. The purpose of the study is, besides assessing TRACE against BETHSY 9.1b test data, also to compare the results with the RELAP5 computer code results and to determine the code accuracy of both simulations. BETHSY is an integral test facility, which was designed to simulate most PWR accidents of interest, to study accident management procedures and to validate the computer codes. It is a scaled down model of three loop Framatome nuclear power plant with the thermal power 2775 MW. The facility consists of pressure vessel, reactor coolant pumps and piping, heat tracing system, the system for break simulation, instrumentation and the control systems. The selected 9.1b test represents the accident scenario with a 5.08 cm equivalent diameter cold leg break without high-pressure safety injection and with delayed ultimate procedure. After the Fukushima-Daiichi accident, International Atomic Energy Agency (IAEA) and Western European Nuclear Regulators Association (WENRA) listed this type of loss of coolant accident as a design extension conditions (DEC). No DEC safety features for high pressure injection were available in BETHSY 9.1b test. Rather, delayed operator action for secondary system depressurization has been studied. The simulations will be carried out with the latest code versions TRACE V5.0 Patch 5 and RELAP5/MOD3.3 Patch 5. The TRACE input model has been developed by conversion of the standard RELAP5 input model of BETHSY, developed at Jožef Stefan Institute during its participation in the OECD/NEA international standard problem ISP 27. ISP 27 was prepared and lead by the Committee on the Safety of Nuclear Installations (CSNI) of the Nuclear Energy Agency within the Organisation for Economic Co-operation and Development (OECD/NEA).
For accuracy quantification the Fast Fourier Transform Based Method by signal mirroring (FFTBM-SM) and original FFTBM will be used. In the paper the qualitative comparison of the simulated results with the experimental measurements will be presented. Finally, the results of code accuracy will be shown as time dependent values through the whole duration of experiment rather than just for the selected timed windows.

08.09.2020 15:40 Poster Session

Materials in nuclear technology - 1101

Thermal parameters study of neutron-irradiated nanocrystalline silicon carbide (3C-SiC) using DSC and TGA methods

Elchin M. Huseynov

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


Over the past few years, silicon carbide is at the focus of world researchers due to the combination of its attractive physical and chemical properties. Simultaneously, SiC has inherently high rigidity, melting temperature (3000K or more depending on polytype), chemical and physical resistance. SiC as a semiconductor has a functional application area at high temperatures. Nanocrystalline silicon carbide (3C-SiC) particles have been irradiated by neutron flux (2×1013 n•cm-2s-1) up to 5 h at the TRIGA Mark II type research reactor. At the present work, thermal properties of nanocrystalline 3C-SiC are comparatively investigated before and after neutron irradiation at the 300K<T<1300K ranges. Simultaneously, the DSC (Scanning Calorimetry), TGA (Thermogravimetric Analysis) and DTG (Differential Thermogravimetric Analysis) experiments were conducted from 300K up to 1300K. Oxidation mechanism of nanocrystalline 3C-SiC particles have been theoretically and experimentally studied before and after neutron irradiation. The kinetics of mass and heat flux were analyzed at the heating and cooling processes using DSC spectroscopy.
The comparative analysis before and after neutron irradiation revealed that, the nanocrystalline 3C-SiC particles have a very resistant physical property under neutron irradiation. Some effects were observed in the DSC curves before and after the irradiation about at 300K<T<800K ranges. It has found out that neutron flux did not affect the activation energy of the nanocrystalline 3C-SiC particles (the activation energy was close to the typical value of 120kJ/mol). Moreover, TGA and DTG analyses have revealed that the oxidation rate at 300K<800K is almost close to zero. However, a slight amount of oxidation was observed in the nanomaterial in the range of about 800K<T <1300K. The effects with a different mechanism are observed in both kinetics of mass and heat flux after neutron irradiation (heating and cooling processes). The temperature deviations around 800K in both heating and cooling processes can be attributed to the typical Debay temperature (?D ~ 800K) for these materials. However, unlike the case, the Debye temperature for silicon carbide is 1270K in the literature. It is also known that the Debye temperature shifts to the direction of a temperature decrease in many nanomaterials. Probably, Debye temperature for the nanocrystalline 3C-SiC particles shifted to approximately 800K at the nanoscale. Although, more analytical experiments are needed to give a more accurate opinion of Debay temperature [1-2]. According to the calculated value of Gibbs energy, 3C-SiC nanocrystals are spontaneous or more resistant than high temperatures at relatively low temperatures.
1. 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
2. Elchin M. Huseynov, Tural G. Naghiyev, Ulviyya S. Aliyeva "Thermal parameters investigation of neutron-irradiated nanocrystalline silicon carbide (3C-SiC) using DTA, TGA and DTG methods" Physica B: Condensed Matter 577, 411788, 2020

08.09.2020 15:40 Poster Session

Materials in nuclear technology - 1102

3D reconstruction of free-surface grain boundaries from 2D EBSD measurement maps

Timon Mede, Samir El Shawish

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


To evaluate local mechanical stresses in a realistic material, a 3D microstructure information of the sample is usually necessary. In particular, the knowledge of grain boundary (GB) locations and GB normal directions is required in order to estimate intergranular normal stresses, which seem to be key ingredient for accurate prediction of intergranular cracking.

In this study, a novel method for 3D reconstruction of free-surface GBs of a polycrystalline metal is proposed in a form of Mathematica script. Only GBs on a free surface of a metal are considered although the methodology presented here can be used at any location along the thickness of the sample. The employed method reconstructs GBs from several 2D image maps stacked atop each other to form a section of a 3D aggregate model. The 2D images can be either synthetic or produced from a realistic material by, e.g., cyclic removal of the parallel layers of the sample, followed by Electron BackScatter Diffraction (EBSD) imaging of the planar sections. In either case, each coloured voxel forming a regular grid of a 2D image represents a local crystallographic orientation. Once the input 2D EBSD image maps are available (here kindly provided by CEA, France), the method identifies first the GB locations and in-plane GB slopes by analysing local orientation contrast within the top horizontal EBSD plane. For each GB voxel identified on a top plane several vertical cross-section planes are then formed and same-coloured voxels identified within each plane in order to calculate the average out-of-plane GB tilt. Once the in-plane and out-of-plane GB slopes are known a 3D GB normal is finally calculated and assigned to the corresponding free-surface GB voxel.

The accuracy of the method is verified against synthetic 2D image maps produced from a 3D Voronoi aggregate model where GB normals are exactly known. A very good agreement for the in-plane GB slopes is found, however, slightly worse reconstruction results are obtained for the out-of-plane GB components. Generally, the accuracy of the method improves with the increasing distance between the considered 2D image planes.

08.09.2020 15:40 Poster Session

Materials in nuclear technology - 1103

The influence of hydrogen on Zircaloy-4 microstructure studied by SEM and nanoindentation after simulating LOCA

Petra Gavelova1, Patricie Halodová1, Daniela Marušáková1, Ondřej Libera1, Věrá Vrtílková2, Jakub Krejčí2

1Research centre Rez, Hlavni 130, 250 68 Husinec-Řež, Czech Republic

2UJP Praha a.s., Nad Kaminkou 1345, 156 10 Praha-Zbraslav, Czech Republic


Zirconium-based alloys play a vital role in thermal-neutron reactor systems, as they represent a first containment barrier to fission products. Thereby the cladding mechanical integrity is crucial for nuclear safety and becomes important at Loss of Coolant Accident, reaching critical temperatures up to around 1200°C. Cladding materials are improved and tested in critical reactor conditions, more then after Fukushima disaster. Zircaloy-4 cladding tubes used for west PWR are examined in our study. The specimens were high-temperature oxidized in steam at the series of temperatures from 950 up to 1200°C to simulate PWR reaching LOCA conditions. To observe the effect of hydrogen in the microstructure, the hydrogenated specimens with ~1000 ppm H were compared to specimens with almost no hydrogen content. SEM-WDS with complementary nano-indentation method were used to characterize Zircaloy-4 microstructure formed after simulating LOCA. Oxygen concentration and nanoindentation profiles performed at the same lines showed the effect of hydrogen on the Zircaloy-4 microstructure. The specimens with 1000 ppm H showed mostly the higher O-content depending locally on the microstructure. A fluctuation of oxygen values in adjacent grains can be caused by preferential oxidation through the favorably oriented oxygen-rich ?-Zr(O) phases studied by WDS and EBSD methods. The specimens with high H-content showed the material hardening with significant difference in ß-phase. It is obvious, that hydrogen has an indirect effect on cladding embrittlement by icreasing of oxygen solubility in zirconium matrix.

08.09.2020 15:40 Poster Session

Materials in nuclear technology - 1104

Development of a T-junction structural model in Abaqus for efficient thermo-mechanical analyses using detailed CFD data

Oriol Costa Garrido1, Nejc Kromar2, Samir El Shawish3, Leon Cizelj3

1Institut "Jožef Stefan", Jamova cesta 39, 1000 LJUBLJANA, Slovenia

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

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


Detailed computational fluid dynamics (CFD) simulations of various fluid phenomena are possible in recent years due to the increase of computational power and the development of dedicated computer codes. One example of fluid phenomena is the turbulent mixing of fluids at different temperatures in T-junctions, which is nowadays resolved by CFD codes while capturing simultaneously the temperatures in the structures surrounding the fluid, i.e., with the so-called conjugate heat transfer (CHT) simulations. The drawback of such simulations is the very stringent mesh requirements in the fluid and structure. The temperatures in the structure, once calculated, are typically used in the subsequent thermo-mechanical analyses to anticipate the strains and stresses. In order to facilitate the transfer of structural temperatures between the CFD and finite element (FE) codes, the mechanical analyses typically employ the same meshes used in the detailed fluid (CFD-CHT) simulation. However, this strategy may increase substantially the computational costs of the mechanical analyses, which is not strictly necessary from the structural mechanics point of view. In the mechanical analyses, denser meshes may be necessary at locations of interest and/or structural discontinuities while rather coarse meshes may suffice elsewhere in the structure.
In this paper, a strategy using Abaqus and python scripting is presented to use very detailed temperature data from a CFD-CHT simulation in the thermo-mechanical analyses of a T-junction with arbitrary FE meshes. The strategy uses the Abaqus-CAE mapped field interpolation and the data output by Abaqus is then post-processed to create the needed input files for the FE simulations. The usefulness of this strategy is demonstrated in a mesh sensitivity analysis with the aim to obtain accurate stresses while minimizing the computational resources in terms of simulation time and disk space. Additionally, the paper also presents the basic analyses of the T-junction with pressure load and steady-state temperature distribution to analyze the complex stress-state in T-junction configurations.

08.09.2020 15:40 Poster Session

Materials in nuclear technology - 1105

Fuel performance modelling at extended burn-up using FEMAXI-6 code

Jakub Lüley, Branislav Vrban, Stefan Cerba, Filip Osuský, Vladimir Nečas

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


The scope of current research in the field of fuel performance is primary aimed to an improvement of the operating reliability, safety and cost effectiveness of the reactors in operation. The current requirement of nuclear industry is to have fuel suitable for load follow operation. Fission gas release, PCMI and stress corrosion cracking are the main phenomena that limit the variability of reactor operation from a safety perspective. To reasonable predict the fuel performance limits it is necessary to benchmark the computational tools against high quality experimental data. This work is devoted to the calculation of fuel performance using the code FEMAXI-6 based on the longest irradiation experiment in a Halden reactor. The fuel burn-up was approaching 90 MWd/kgUO2 in three selected rods which were equipped by the pressure sensors and were subjected to extensive post-irradiation examination. During the experiment, the rods were exposed to several periods of power cycling. The rods were manufactured with different fuel grain size and fuel-to-clad gap size. Information about rods diameters before and after irradiation are available and hence the dimensional changes can be evaluated. Depletion calculation is also carried out using SCALE / TRITON sequence to compare prediction of the local burn-up and rare gases production.

08.09.2020 15:40 Poster Session

Materials in nuclear technology - 1106

Microstructural characterization of the early-stage swelling in He implanted ferritic/martensitic steels

Vladimir Krsjak1, Jarmila Degmova1, Stanislav Sojak2, Vladimír Slugeň2

1Slovak University of Technology Faculty of Electrical Engineering and Information Technology Department of Nuclear Physics and Technology, Ilkovičova 1, 812 19 Bratislava, Slovakia

2Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia


Ferritic/martensitic (f/m) steels are candidate structural materials for nuclear applications due to their radiation resistance and void-swelling resistance. In the radiation environments with a high production rate of helium, such as fusion or spallation applications, these materials suffer from a non-negligible swelling due to inhibited recombination between vacancy and interstitial-type defects. In this work, the swelling of helium implanted Fe9Cr steel and its oxide dispersion strengthened (ODS) variant have been investigated by means of slow positron beam spectroscopy and transmission electron microscopy. The void swelling behaviour and radiation-induced open-volume defects were characterized in a wide-scale depth profile quantitative analysis. The obtained data show an onset of the incubation stage in the f/m steel strengthened by dispersion of oxide particles followed by a non-steady state swelling. This onset is determined by the pre-existing concentration of vacancy-type defects defect, which was found to be higher in the ODS steels compared to conventional f/m steel. These vacancy-decorated nanoparticles act as sinks for radiation-induced defects and helium and are responsible for an improved swelling resistance of the ODS steels. At high dpa and high helium concentration, however, the swelling of both studied materials saturate at roughly the same value. This suggests that the performance of the ODS steel in extreme radiation environments with a high production rate of helium does not overcome the performance of conventional f/m steels.

08.09.2020 15:40 Poster Session

Materials in nuclear technology - 1107

Automated Prefabrication Welding and Advanced Ultrasonic Testing of ASME III Div.1 – Subsec. NB Class Piping Welds

Joško Andrejčič1, Matej Pleterski1, Jernej Jerman2, Rok Topolnik3, Damjan Klobčar4

2Q Techna d.o.o., Cvetkova ulica 27, 1000 Ljubljana, Slovenia

3Q-Techna, Knezov štradon 92, 1000 LJUBLJANA, Slovenia

4Faculty of Mechanical Engineering, University of Ljubljana, Aškerčeva 6, 1000 Ljubljana, Slovenia


Nuclear safety-related piping systems play an important role in the safe operation of nuclear power plants. For safety Class 1 piping systems, the fabrication demands are most stringent and include special requirements for welding and examination of welds. This paper presents automated welding in prefabrication of safety-related, heavy wall austenitic stainless steel piping, utilizing combination of Gas Tungsten Arc (GTAW) and Gas Metal Arc (GMAW) welding processes. Further on, welds were subjected to destructive and non-destructive examination (NDE). NDE has been performed by advanced ultrasonic testing (UT) techniques by encoded scanning, utilizing combined Phased Array (PAUT) and Time of Flight Diffraction (TOFD) techniques. In such case, ultrasonic testing is applied in lieu of radiography in accordance with ASME Code. Ultrasonic testing serves for the purpose of volumetric examination to detect fabrication flaws and Pre-Service Inspection (PSI). Compared to manual, automated combined process welding exhibits major increase in productivity, weld quality reproducibility, and improvement of weld root geometry characteristics.

08.09.2020 15:40 Poster Session

Regulatory issues and legislation - 1401

SNSA's methodology for inspection supervision of the Krško NPP refuelling outages

Sebastjan Šavli, Matjaž Podjavoršek, Matjaž Pristavec, Aleš Janežič, Tom Bajcar

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


The SNSA only has a small number of nuclear inspectors who, despite the fact that Slovenia operates only one NPP, have to inspect the same extent of the NPP's safety areas as the large nuclear countries do. While the large regulators may employ, in addition to general and resident inspectors, inspectors specialists covering in detail their particular areas of interest, four SNSA’s nuclear inspectors have to supervise wide range of the NPP's operation on their own. The SNSA’s inspectors have a high level of knowledge and experience regarding the NPP as a whole, so they are able to assess the interactions of the activities from different disciplines and quickly gain a large picture.

However, for such a small inspection team it is a serious challenge to carry out in-depth, detailed and comprehensive inspections for each individual operational area. Therefore, during the Krško NPP power operation inspectors often include experts from the SNSA's Nuclear Safety (licensing) Division into the inspection teams. These SME's (subject matter experts) provide support to inspectors during thematic inspections in their areas of expertise.

A special inspection approach and organisation is being implemented to effectively supervise a large amount of activities performed by the Krško NPP staff and by the contractors within a limited time frame during refuelling outages. Through the effective inspection supervision the SNSA seeks to verify and ensure that all outage activities are prepared and implemented in accordance with the regulation in force, the standardisation, best engineering practice and a high level of safety culture. All this ensures that the radiation and nuclear safety are maintained on a high level at all times and that the NPP is ready for a safe operation during the next 18-month fuel cycle as well as for a safe long term operation.

Unlike during the power operation, when typically one inspection per week is performed, during the refuelling outage the SNSA applies extended inspection supervision by exercising a continuous on-site presence of its inspectors as well as with periodic thematic visits of other SNSA's experts to the plant. Furthermore, authorised technical support organizations (TSOs) perform in-detail oversight of the selected outage activities, each in their area of expertise. Unlike the inspectors the TSOs do not have enforcement powers and cannot require any corrective actions to be taken by the NPP. However, they regularly report their findings to inspectors who address the TSO's findings and recommendations later on during special inspection reviews.

The provision of support by the TSOs is a specific feature of the SNSA's methodology for performing its outage inspections, which gives the inspectors a deeper insight into the preparation and implementation of the outage activities. However, the SNSA faces different challenges regarding the scope and content of the TSO's oversight, the requested regulation provisions, as well as the control of their work and avoidance of the conflicts of interest. The consent for the reactor criticality and the consent for the transition to the power operation are given by the TSOs based on their oversight during the outage. This has been practiced for many years, however there is a lack of suitable requirements for such practice in the existing regulations. Therefore, improvement of the existing regulations by adding provisions regulating the role of TSOs during the NPP's outages is one of SNSA's challenges. Coping with challenges mentioned above should lead to further improvements which will make the regulatory inspection of the Krško NPP outages more predictive, transparent and effective.

08.09.2020 15:40 Poster Session

Regulatory issues and legislation - 1402

Development of the Severe Accident Analysis Capabilities at the Slovenian Nuclear Safety Administration

Tomi Živko, Andreja Peršič, Tomaž Nemec

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


The Slovenian legislation requires that the facility operator of a radiation or a nuclear facility shall submit to the Slovenian Nuclear Safety Administration (SNSA) an application for the approval of significant modification which shall be supplemented by an expert opinion of an authorized expert on radiation and nuclear safety.
In many cases authorized experts must evaluate the impact of proposed modification on the nuclear safety and for this purpose they use computer codes for simulation of accidents at nuclear facility. At the turn of century, the only Slovenian nuclear power plant, the NPP Krško, replaced its steam generators and uprated the plant power. The SNSA decided to use the MELCOR code to check effects of the new plant configuration in case of severe accidents. MELCOR is a computer code that models the progression of severe accidents in light water reactor nuclear power plants. It was developed for the United States Nuclear Regulatory Commission (NRC). An authorized expert developed a model of the Krško NPP for the MELCOR. MELCOR was used at the SNSA as an independent analysis tool for severe accidents.

Due to other urgent priorities and lack of manpower, usage of MELCOR at the SNSA stopped. After the Fukushima accident in 2011, the Krško NPP started the Safety Upgrade Program that introduced new safety features according to the Design Extension Conditions (DEC). For this Beyond Design Basis Accidents (BDBA) domain, the severe accident scenarios were analyzed with the Modular Accident Analysis Program (MAAP) by the plant designer and as an independent code, MELCOR was used by the authorized expert. In 2015 Slovenia renewed cooperation in the US NRC severe accident research programme CSARP (Cooperative Severe Accident Research Program). Membership in the CSARP programme enables usage of the computer code MELCOR. In last years the SNSA sponsored several research and development projects with goal to evaluate the impact of proposed modification of the Krško NPP to Severe Accident Management Guidelines (SAMG). Analyses were performed using new revision of MELCOR code. Adequacy of the existing SAMG strategies was confirmed. This new version of MELCOR code was installed at the SNSA. The foreseen goal of MELCOR usage at the SNSA is to improve technical competence of employees, evaluate impact of proposed modifications to safety of the plant and estimation of source terms in case of severe accidents. Development of the MELCOR model of the NPP Krško is also presented. The analyses performed with MELCOR code and the development of the Krško NPP model resulted in improved competence of the authorized expert in the area of severe accidents and enabled development of the Severe Accident Analysis Capabilities of the SNSA, which includes understanding of severe accident phenomena in the containment.

08.09.2020 15:40 Poster Session

Regulatory issues and legislation - 1403

Post-Accident Strategy after the Nuclear or Radiological Accident and Lessons Learned from COVID-19 Crisis

Helena Janžekovič

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


In 2014 the Slovenian Nuclear Safety Administration (SNSA) developed the post-accident strategy presented in the document Post-Accident Strategy after the Nuclear or Radiological Accident (the Strategy). This was the very first attempt to establish such strategy in the State. The Strategy has been presented internationally among others at the IAEA and NENE 2015. It was noted at that time that such national strategies have been prepared rarely. Very limited international experiences have been published. The preparation of such strategy requires strong cooperation of many stakeholders, e.g. in France the Steering Committee on Post-Accident Management was established in 2005.
The development of the SNSA Strategy was initiated by two facts, namely:
1. Fukushima accident in 2011 which posed a huge challenge to one of the most developed countries in the world, i.e. Japan.;
2. European Union (EU) legislation, i.e. Council Directive 2013/59/EURATOM, which requires in Article 98 that “the emergency response plans shall also include provision for the transition from an emergency exposure situation to an existing exposure situation”.
Furthermore, in 2018 European Commission organized the scientific seminar on “Management of long-term exposure after a nuclear or radiological accident” which stressed among others the importance of »local reference levels« for the management of long-term exposure, holistic evaluation of the situation and stakeholder involvement.
While it is well known that nuclear and radiological accidents as such might challenge any country it is less known that post-accident management might be an even bigger challenge for decades when handling public health and contaminated areas just to mention two issues. As an example, managing fires on contaminated areas might be a particular challenge. It must be noted that international aspect of accidents might be taken into account for decades, e.g. in EU particular food restrictions are valid even 30 years after the Chernobyl accident.
The Strategy mentioned took lessons learned not only from nuclear and radiological accidents but also from other accidents. In 2020 the COVID-19 crisis is posing a huge challenge to all countries. On the other hand, it offers opportunity to better understand how to manage such huge crisis in the State and internationally. In light of the lessons learned from COVID-19 crisis the article is critically reviewing the SNSA Strategy. Proposal for its upgrading are discussed.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1501

Modelling of coupled hydraulic-mechanical behaviour of MX-80 bentonite under hydration with groundwater

Asta Narkuniene, Darius Justinavicius, Povilas Poskas

Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas, Lithuania


As stated in the EU directive 2011/70/EURATOM [1] every member state (country) is responsible for the implementation of a safe and sustainable solution for SNF and radioactive waste management and disposal. Radioactive waste disposal is among the basic radioactive waste management steps following the nuclear reactor operation and decommissioning. Geological disposal concept is based on multibarrier system made of engineered and natural barriers. The radioactive waste management development program has been approved by Lithuanian Government in 2015. It foresees the geological disposal of RBMK-1500 spent nuclear fuel from Ignalina nuclear power plant (Lithuania) starting in 2066.
Particular clay (bentonite) is among the materials considered for engineered barriers due to its properties such as low hydraulic conductivity, high sorption capacity, swelling potential, etc. This paper presents the results of the analysis on coupled hydraulic-mechanic MX-80 bentonite behaviour under its hydration with groundwater. Numerical model was based on Richard’s equation and linear swelling model including non-linear couplings. Numerical modelling was performed with computer code COMSOL Multiphysics (USA) and the results were compared to experimental results of hydration of MX-80 bentonite sample at laboratory. The overall modelling results were in line with experimental data. Data on measured void ratio, dry density, moisture content at different parts of specimen were slightly underestimated or overestimated with more significant differences for upper part of specimen. Modelled axial pressure was estimated being higher than radial pressure while measured radial pressure was higher than the axial pressure in the experiment.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1502

Neutron absorber for VVER-1000 final disposal cask

Martin Lovecky1, Jiri Zavorka1, Jana Jirickova2, Radek Skoda3

1University of West Bohemia, Univerzitni 8, 306 14 Plzeň, Czech Republic

3Czech Technical University, Zikova 1903/4, 166 36 Prague 6, Czech Republic


The recent increasing demand for better nuclear fuel utilization requires higher enriched uranium fuels which is a challenge for spent fuel handling facilities in all countries with nuclear power plants. The operation with higher enriched fuels leads to reduced reserves to legislative and safety limits of spent fuel transport, storage and final disposal facilities. In some cases, the required boron amount in the absorber plates or tubes can be higher than current metallurgy processes allows. This study addresses the neutron absorber solution with significantly increased nuclear safety and improved economics where a new concept of inseparable neutron absorber is introduced to achieve fuel reactivity decrease. Same or better criticality safety is achieved with significantly lower or even no boron content in the cask basket absorber. Alternatively, it is possible to reduce fuel assembly pitch with the same boron amount and subsequently decrease overall cask dimensions and its cost. Because of less strict requirements for absorber material when compared to active core environment and better spatial position inside spent fuel handling facility, the choice of absorber material expands currently used boron element. Erbium, cadmium, gadolinium, hafnium, samaria and dysprosium elements are among the most suitable materials. Criticality safety analysis of the recent VVER-1000 final disposal cask with 3 fuel assemblies is performed with the new neutron absorber concept.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1503

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

Riku Tuominen1, Ville Valtavirta2

1VTT Technical Research Centre of Finland Ltd, P.O. Box 1000, FI-02044 VTT, Finland

2VTT Technical Research Centre of Finland Ltd., P.O. Box 1000, FI-02044 VTT, Finland


The estimation of spent nuclear fuel source term (decay heat, reactivity, nuclide inventory etc.) has several sources of uncertainty such as uncertainties in nuclear data, uncertainties in the operation history, choice of calculation parameters etc. In this work the effect of calculation parameters is studied by estimating the source term with the built-in burnup capability of Serpent 2. The effect of following parameters is considered: depletion zone division, burnup steps, unresolved resonance probability table sampling, Doppler-Broadening Rejection Correction (DBRC) and energy dependent branching ratios. As a test case a 2D BWR fuel assembly was modelled. First a burnup calculation was executed up to a burnup of 60 MWd/kgU followed by a decay calculation up to 10^7 years starting from the maximum burnup. These simulations were repeated for different variations of the studied parameters and since the use of Monte Carlo method introduces statistical uncertainty, each variation was repeated five times to obtain an estimate for the uncertainty. The following source term components were considered when investigating the effect of the studied parameters: total decay heat, photon emission rate and spontaneous fission rate. In general the differences resulting from the use of different parameter variations were small for all three studied source term components. For the decay heat largest absolute relative difference was approximately 0.6 % and for the photon emission rate approximately 1.1 %. For the spontaneous fission rate maximum absolute relative difference of nearly 8 % was observed. For all three components the variation of the depletion zone division resulted in the largest relative differences. Clear differences were also observed for burnup step length and DBRC variations. The use of unresolved resonance probability table sampling and energy dependent branching rations had an insignificant effect on the studied source term components.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1504

Development of Neutron Source for Nuclear Power Plant Decommissioning

Kristyna Gincelova1, Martin Lovecky2, Radek Skoda3

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

2University of West Bohemia, Univerzitni 8, 306 14 Plzeň, Czech Republic


During the decommissioning of nuclear power plants, one of the most important operation is appropriate classification of radioactive waste. Radioactive waste from the operation covers a wide range from low-level waste to high-level waste. With exception of spent nuclear fuel, significant part of radioactive waste covers the components of primary circuit, mainly the reactor pressure vessel and its components. For nuclear power plants decommissioning the waste management optimization is essential. In case of primary circuit the optimization is based on proper determination of its components activities. This paper presents the neutron source development that is the first step of waste management optimization process. Presented neutron source was derived from data set covering the 30 years operation of Dukovany Nuclear Power Plant.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1505

Impact of approximations in operating history data on spent fuel properties with Serpent 2

Silja Häkkinen


Accurate knowledge of spent nuclear fuel (SNF) properties is important when planning final disposal, handling and intermediate storage of SNF. Decay heat and reactivity determine how densely the SNF canisters can be placed in the repository tunnels [1]. Activity must be considered in the handling and intermediate storage for appropriate radiation shielding. Nuclide inventories are also needed for many purposes such as reactivity calculations and dose estimates from nuclides propagating to the biosphere such as C-14, Cl-36, I-129, Mo-93 and Ag-108m [2]. More accurate knowledge of SNF properties and related uncertainties yield cost savings due to reduced margins in e.g. repository space.

Computational characterization of SNF involves many uncertainty sources one of which is uncertainty in operating history. Uncertainties can be related e.g. in the accurate knowledge of operating history parameters such as fuel and coolant temperatures, reactor pressure, power density and boron concentration or in the intentional approximation of these parameters in order to simplify the calculation. For example, the accurate modelling of a fuel assembly’s operating history spanning over three or four years using a Monte Carlo code is not always considered practical and some approximations can be made.

In this work, the operating history of VVER-440 fuel assemblies based on weekly monitoring data are modelled accurately using the Monte Carlo particle transport code Serpent 2 [3]. Then the monitored parameters such as coolant temperature and density, boron concentration and assembly power are averaged one by one over the simulated operating history. The effect of the averaging on important spent fuel properties such as e.g. decay heat, activity and nuclide inventory is investigated. The purpose of the calculations is to determine how rough approximations in the operating history of the investigated parameters can be made without significant effect on spent fuel properties and to determine which of the investigated parameters are most sensitive to approximations.

1. Ikonen, K.,, “Thermal Dimensioning of the Olkiluoto Repository - 2018 Update”, POSIVA Working Report 2018-26, 2018.
2. Posiva Oy, “Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto - Synthesis 2012”, POSIVA 2012-12, 2012.
3. Leppänen, J., et al. (2015) "The Serpent Monte Carlo code: Status, development and applications in 2013." Ann. Nucl. Energy, 82 (2015) 142-150.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1506

Evaluating the effect of decay and fission yield data uncertainty on BWR spent nuclear fuel source term

Antti Rintala

VTT Technical Research Centre of Finland Ltd, P.O. Box 1000, FI-02044 VTT, Finland


Knowledge of spent nuclear fuel (SNF) source term, for example decay heat, reactivity and nuclide inventory of SNF, is essential in the safe handling and final disposal of SNF. The computational characterization of SNF has many sources of uncertainties. One of these is the uncertainties present in nuclear data. This work continues the identification of the major uncertainty components in the decay and fission yield data in relation to the SNF source term.

The uncertainties of decay and fission yield data are propagated from ENDF-6 file format nuclear data library in the SNF source term using a sampling based technique in the Monte Carlo particle transport program Serpent 2. Studied components of the SNF source term are produced multiple times as functions of burnup and cooling time with randomly perturbed decay and fission yield data.

Similar work has been previously conducted for VVER-440 type fuel assembly. In this work, a BWR type fuel assembly is studied. The study is conducted varying the void fraction to examine its effect on the uncertainties of the SNF source term components caused by the uncertainties in the decay and fission yield data. Other nuclear data related uncertainties are ignored in this study, and only fixed, nominal depletion conditions are considered.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1507

Uncertainty of radioactive waste characterization using calorimetry combined with gamma spectrometry – numerical case study

Wojciech Kubiński1, Cedric Carasco2, Daniel Kikola3, Christophe Mathonat3, Denise Ricard4, Dariusz Tefelski3, Holger Tietze-Jaensch5

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

2CEA Cadarache, IPSN/DRS/SEMAR BAT. 700, 13108 St. Paul Lez Durance, France

4ANDRA, 1 a 7, rue Jean Monnet Parc de la Croix Blanche, 92298 Chatenay Malabry Cedex, France

5European Spallation Source ERIC, Box 176, S-221 00 Lund, Sweden


To determine the content of multi-nuclide waste drums, gamma spectrometry combined with calorimetric measurements is often used to reduce the measurement uncertainty arising from the presence of radioactive elements shielded or deeply buried inside the drum or emitting non-penetrative radiation to (e.g. pure beta-emitters like Sr/Y-90) [1,2]. For such materials, part of the radiation may not be able to leave the drum and cannot be detected by standard methods. Calorimetry, however, can potentially detect all of the radiation as the energy of the deposited particles eventually turns into heat [2]. On the one hand, the gamma radiation coming out of the drum allows to partly determine the composition of the sample by measuring the emitted spectrum. On the other hand, the part of the radiation that stays inside the drum (alphas, betas and low-energy gammas) can be determined by calorimetric measurements. Thus, characterization based on these two methods can lead to a reduction in the final measurement uncertainty [3,4].
One of the goals of the CHANCE-H2020 project, under which this study was carried out, is to use calorimetric measurements to reduce the uncertainty of waste characterization [3,4]. In the frame of the project, a series of simulations were carried out using the Geant4 code. Four typical compositions of the waste drum were considered. It was studied what spectrum can be registered by gamma spectrometry and what amount of heat can be measured by calorimetry. Then, it was calculated what would be the isotopic composition of the radioactive elements determined using the combination of both measurements. In this way, four cases were analysed, determining the final measurement uncertainty, showing, among others, that some of the isotopes (especially placed deep in the matrix), can be invisible to gamma measurements but detectable by the combination of spectrometry and calorimetric assay.

[1] D.S. Bracken, C.R. Rudy, Principles and applications of calorimetric assay, LA-UR-07-5226
[2] D.S. Bracken, R.S. Biddle et. al., Application Guide to Safeguards Calorimetry, LA-13867-M, January 2002
[3] D. Ricard, S., Plumeri, L. Boucher, A. Rizzo, H. Tietze-Jaensch, C. Mathonat, C. Bruggeman, J. Velthuis, L Thompson, G. Genoud, C. Bucur, D. Kikola, G. Zakrzewska-Koltuniewicz, (2018). The CHANCE project “Characterization of conditioned nuclear waste for its safe disposal in Europe”. DEM 2018 - Dismantling Challenges: Industrial Reality, Prospects and Feedback Experience, France, Avignon – 2018, October 22 – 24
[4] CHANCE 2020 :

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1508

Development of MARSSIM WRS Test Verification Tool

Ara Go, Jungjoon Lee, Kyungwoo Choi


In Korea, The Commission shall inspect whether the final site status report satisfies the criteria for reuse of the site and the remaining buildings as determined and published by the commission in accordance with Article 23-5 of the Enforcement Regulations of the Nuclear Safety Act. In other words, if a licensee applies MARSSIM (Multi-Agency Radiation Survey and Site Investigation Manual), the most widely used for investigating the residual radioactivity of sites at decommissioning completion stage, the regulatory body should review whether the final site status survey conducted in accordance with MARSSIM is appropriate. Among them, licensee submitted measurement data by survey unit and confirmation of the suitability of statistical analysis is a part that does not require regulatory judgment and can be performed quickly and accurately if programmed. Therefore, the object of this study is to develop MARSSIM Wilcoxon Rank Sum (WRS) Test verification tool. WRS Test is a non-parametric statistical methodology that is used when the nuclide is in background at a significant fraction of the DCGL. To develop WRS Test verification tool, WRS Test procedure proposed by MARSSIM was programmed using Excel and R. The WRS Test procedure was divided into 4 steps, and in order to run the program, DGCL, survey unit and Type? and ? error must be determined first. In the case of Excel, even if user don’t have any statistical knowledge or experience in using statistical programs, the results can be obtained by evaluating them according to the procedures given in Excel. However, there is the inconvenience that some information cannot be loaded automatically and the user must input it manually. R, a commercial statistical program, allows users to read multiple input files for each survey unit at once and make statistical processing if the user creates the input files according to the format. It also provides calculation results in Excel and graphs. However, there is a disadvantage that basic knowledge of the R program is required. In order to check whether the developed verification tool works properly, six sample examples were created for each case. Through the example data, it was confirmed that the verification tool is operating normally, which can save time and manpower required for data verification and statistical processing. The verification tool developed in this study can be used to verify the results of the WRS Test at the decommissioning completion stage.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1509

Sign Test Verification Tool of Investigation Result for Residual Radioactivity of Building and Site at Decommissioning Completion Stage of Nuclear Facilities

Jungjoon Lee, Ara Go


Kori unit 1, 1st commercial power reactor in Korea, has been permanently shutdowned in July 2017. KHNP, licensee as an operator of Kori unit 1 is now preparing the application of decommissioning approval according to the Nuclear Safety Act. Final decommissioning plan should include the manual of survey and investigation for residual radioactivity of building and site in the stage of decommissioning completion.
Several countries are adapting MARSSIM for building and site during or after decommissioning of facilities. MARSSIM mainly suggests non-parametric statistical sampling method which consist of Sign test.
Regulation authority review final site survey report when licensee submit decommissioning completion report. It would be main issue on the analysis of radiation survey and site investigation result. This study shows verification tool of investigation result for residual radioactivity of building and site at decommissioning completion stage of nuclear facilities. Verification is available on MS excel and R program. Several simulation test using manufactured data for Sign test shows the tool could be useful and contribute to review process making it easier and faster.

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1510

Substantiation of the need to remove fuel-containing materials and development of the concept of technological solutions for the transformation of Shelter Object into an environmentally safe system

Serhii Kupriianchuk, Serhii Paskevych, Volodymyr Rudko

Institute for Safety Problems of NPP’s, National Academy of Sciences of Ukraine, 36a Kirova str., Chornobyl, Kyiv reg.,Ukraine, 07270,, Ukraine


The Shelter object contains tens of thousands of tons of radioactive waste (RW), including a significant amount of high-level waste (HLRW) and fuel-containing materials (FCM). For the present day, the final number of FCM and HLRW has not been determined. So the amount of fuel on the upper marks of SO is, according to various estimates, in the range from 20 to 110 tons (U). There is no access to many FCM clusters, and the presence of nuclear clusters is not ruled out. This is evidenced by the anomalous increase (more than 60 times) in the neutron flux in the SO, which was in June 1990, its retention at a high level, and the decline recorded at the end.
The nature of the placement of FCM inside the SO and their properties have been the subject of systematic experimental study since the creation of the SO. During this time, the main accumulations of FCM were identified and their physical-mechanical, thermophysical, magnetic, some optical and structural properties were studied. However, practically no work was carried out on the development of technological solutions for the removal of FCM from the SO, future treatment of them, including placement in temporary storage with subsequent disposal. Visual observations indicate that the destruction of FCM arrays, their constant fragmentation, with a decrease in the size of the fragments, which creates a potential risk of formation of highly active dust and increase the surface area of FCM. Several experimental studies and visual observations show that due to embrittlement and erosion from the surfaces of FCM, there is a spontaneous separation of fine particles and their deposition on the surrounding surfaces. Laboratory studies conducted in 2011 will show that the rate of surface wind erosion of brown lava type FCM reached up to 19 µg/(cm2 • year), which is two orders of magnitude higher than the same value obtained twenty years ago.
Therefore, one of the most important tasks in the transformation of the Shelter object into an environmentally safe system, and its decommissioning is to address the issue of the treatment of FCM and HLRW of the Shelter.
The conducted researchers allowed to assess the existing possibilities and to substantiate the basic technological decisions on the extraction of FCM from the Shelter object using the NSC systems.
Based on data on the number and location of FCM, seven specific areas of their extraction were identified using different technologies. The removal of FCM can be divided into three stages: priority extraction zones (zones 1-2); area where work with FCM can be carried out without NSC infrastructure (zone 3); and areas where it is hypothetically possible to bury FCM in the place(zones 4-6).

08.09.2020 15:40 Poster Session

Fuel cycle, radioactive waste, and decommissioning - 1511

Simulation of radiation characteristics of fuel-containing materials of the Shelter object of Chernobyl NPP

Serhii Kupriianchuk1, Anatoliy Doroshenko1, Alexsandr Mikhailov1, Andrey Sizov1, Marjan Kromar2

1Institute for Safety Problems of NPP’s, National Academy of Sciences of Ukraine, 36a Kirova str., Chornobyl, Kyiv reg.,Ukraine, 07270,, Ukraine

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


The main radiation hazard of the Shelter object is coming from fuel-containing materials (FCM), which due to their condition and composition, are long-lived radwaste. In the future, within the framework of the strategy transformation of Shelter object and the program of transformation of the Shelter object into an environmentally safe system, activities on extraction of fuel-containing materials are planned.
The objective of this work was to perform a shielding analysis of the FCM stored in a protective container, taking into account different structures of the FCM, with the aim to select the optimal design of a container for transportation and storage.
FCM is divided into several different types: black ceramics, brown ceramics, granular FCM, pumice. They differ in the quantity of fuel, chemical structure, materials composition, and porosity. However, a clear structure and exact characterization of each type is completely unknown.
To develop the FCM composition model for the analysis of dose rates, calculations of the isotopic composition for spent nuclear fuel RBMK - 1000 were previously performed using the SCALE software code. Calculation sequence included TRITON-NEWT modules for fuel burnup and ORIGEN for fuel cooling after operation. Beside isotopic composition, calculations provided neutron and photon energy spectrum and intensity.
The main experimental data of material analysis, which were conducted at the Institute for Safety Problems of NPP of the National Academy of Sciences of Ukraine, are considered in the material balance. Data show that the structure of different FCM types mainly differs in the amount of the fuel and changes in porosity.
In the initial calculation of dose rates, a MCNP model (code verion 6.2 was used) of the FCM, simulating the so-called "black ceramics", was developed. According to experimental data, it was determined that the amount of U fuel is 53%, and the porosity of the structure is 26.1% of the total mass. Stainless steel or lead was used as a container protective material. The total weight of the prototype container with the source was limited to 2 tons. Container was modeled as a cylinder containing FCM with different energy and intensity source simulating the spent RBMK-1000 fuel at different cooling times.
A sensitivity analysis was performed taking into account different porosity of the FCM (starting from 10% with a steps of 10% to 60%), covering several material compositions, which basically results in different amount of the fuel in the matrices of FCM. Dose rate calculations were performed at several radial points (at the surface of the protective container and at points 10 cm and 100 cm from the surface of the container). In addition, dose rate sensitivity of the thickness of the protective shielding layer was also performed Also, simulating the source of FCM as a point source, and the composition of FCM as a protective material, we obtained the value of the effect of self-absorption of dose rate by the composition of FCM
The results of this work will be used for the design of a protective container developed for the transportation and storage of the FCM after their removal.

08.09.2020 15:40 Poster Session

Education and training - 1601

The possibility for distant training and examination in radiation protection in Slovenia

Matjaž Koželj, Vesna Slapar Borišek

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


COVID-19 pandemic has influenced different aspects of our life and work, education of children, adolescents and young people, as well as different training activities of people of all ages related their hobbies, interests and also occupational obligations and aspirations. While certain forms of work could be successfully implemented from “home office”, other forms either require certain tools or equipment or are a part of some chained operations and could not be successfully extracted from the production process. As far as education is considered, we have seen that transformation to distant forms of learning is possible using the existing informational infrastructure. Since this transformation has happened practically overnight and teachers and students were unprepared, we can hardly say that everything went on without problems and mistakes. The most important criterion was that the education process is taking place, and problems should be solved “on the way”. The most important problems to be addressed are practical exercises, evaluation and grading. At the moment, possible problems and complications in these areas are outweighed with the requirement for strict implementation of health-protective measures.
Training in general and especially different forms of occupational training are usually a mixture of lectures, demonstrations and practical exercises. Duration is normally limited (hours, days, weeks) since objectives are precisely defined. This requires more practical involvement from training participants and focused evaluation which is formally required for final certification. Considering the purpose of occupational training, it is not possible simply to readjust (or transform) the process, criteria and evaluation due to some other requirements. This is the reason why many training events were postponed or cancelled and not simply transferred to the internet.
According to current Slovenian legislation, radiation protection training should be implemented in the form of courses, which consist of lectures and practical demonstrations and exercises. Institutions, which organize training and the course programs are approved, as well as authorized radiation protection experts involved in the course implementation. A written examination is required and passing criteria prescribed. Considering all these requirements, it seems that it is not possible to implement some modern forms of distant training and examination. This has been already done in other countries in the past (e. g. in Spain, France, Croatia, etc.) where the training of certain categories of workers is implemented in the form of web courses, and examination is also performed through “electronic forms”.
In our contribution, we would like to discuss the possibility to use these modern forms of training in Slovenia, determine categories and program segments that could be transferred to distant forms of training, and to propose some changes to legislation to enable implementation of this modernized radiation protection training.

08.09.2020 15:40 Poster Session

Education and training - 1602

Distance nuclear training

Tomaž Skobe

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


The paper will present experiences, good practices and feedbacks from distance nuclear training course, that was performed by Nuclear Training Centre during coronavirus pandemic (Covid 19). The Nuclear Training Centre (the acronym in Slovenian language is ICJT), a part of the Jožef Stefan Institute (IJS), started with the training of the nuclear workers at the beginning of commercial use of nuclear technology in Slovenia. The course NPP Technology was started in January 2020, but it was interrupted in the middle of March due to the pandemic circumstances. After that distance training was introduced. The course NPP Technology (the acronym in Slovenian language is TJE) 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 NPP). Approximately 5 months are devoted to different topics, such as nuclear and reactor physics, thermal-hydraulics and heat transfer, radiation protection, electrical engineering, chemistry, materials and nuclear safety.

08.09.2020 15:40 Poster Session

Education and training - 1603

Teaching of reactor physics using a real-time research reactor simulator during COVID-19 induced lockdown

Jan Malec1, Luka Snoj2

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

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


Research nuclear reactors are among others commonly used for education and training of nuclear power
plant (NPP) personnel, research reactor (RR) operators, students of physics, nuclear engineering and related physics.
In 2020, a COVID-19 pandemic has prevented physics students in Slovenia from completing the practical exercises at the Triga Mark II reactor in Podgorica, which are a vital part of the reactor physics education. The Research Reactor Simulator team has adopted the research reactor exercises so they can be carried out on a simulator instead of a physical reactor.

The Research Reactor Simulator, developed at the Jožef Stefan Institute (JSI) simulates time behavior of reactor power, fuel temperature and reactivity by using the 6-group point kinetics equation with feedback.
It features temperature feedback mechanisms as well as xenon poisoning. Physics properties of the simulated reactor can be adjusted to mimic different reactors. Hardware assisted graphics acceleration allows the simulator to display fluent graphics.
The simulator has been under continuous development since its first release in 2017. In 2019, a new thermodynamic model was developed that can qualitatively describe the behavior of Triga-like research reactors in their entire temperature operating range from
\SI{1}{\milli\watt} to \SI{1}{\giga\watt} as a replacement of two thermodynamic models needed to cover the operating temperature range.

The Research Reactor Simulator has more than 240 unique users based on website statistics and is being actively used as a teaching tool at the Faculty of Mathematics and Physics, University of Ljubljana.

08.09.2020 15:40 Poster Session

Public outreach - 1701

Public Opinion about Nuclear Energy – Year 2020 Poll

Radko Istenič1, Igor Jenčič2

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 year 2020 we expect a significantly lower number due to limitations imposed by covid-19 pandemic. 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. For younger visitors, a lecture about energy and an energy workshop is available. The visit includes a demonstration of radioactivity and a guided tour of a permanent exhibition.

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. This year only 353 polls were collected before the covid-19 limitations were imposed. Despite lower statistical accuracy, no major change in public opinion was observed compared to previous years.

08.09.2020 15:40 Poster Session

Severe accidents - 1801

Analysis of Radionuclides Distribution in Krško NPP During Severe Accident

Matjaž Leskovar, Mitja Uršič

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


The response of the Krško NPP containment following a severe accident was analysed with the MELCOR code version 2.2 taking into account mitigation measures for heat removal from the containment solely by the planned 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. Four scenarios were analysed: (1) no mitigation, (2) alternative safety systems available 24 h after initiating event with water injection through containment sprays, (3) alternative safety systems available 24 h after initiating event with water injection into the reactor coolant system, (4) alternative safety systems available 24 h after initiating event with water injection simultaneously through containment sprays and into the reactor coolant system.

The four analysed severe accident scenarios and the applied Krško NPP MELCOR code model are presented. The simulation results of the given scenarios are provided and thoroughly discussed. The main focus is given on the radionuclides distribution and the resulting heat loads. In all considered scenarios, by far the largest amount of residual heat is released in the reactor cavity, comparable in the deposits and water, and the least in the atmosphere. In all mitigated scenarios, on the filters the fraction of the initial radionuclide inventory is similar, but significantly lower than in the unmitigated scenario. The calculations revealed that the filters heat loading strongly depends on the complex chemical processes of the radionuclides.

08.09.2020 15:40 Poster Session

Severe accidents - 1802

Demonstration of the E-BEPU Methodology for LB-LOCA in NPP with PWR Reactor

Piotr Mazgaj1, Piotr Darnowski1, Aleksej Kaszko2, Javier Hortal3, Milorad Dusic3, Rafael Mendizábal4, Fernando Pelayo3

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

2National Centre for Nuclear Research, ul. Andrzeja Sołtana 7, Otwock-Świerk, Poland

3NUCCON, jedrska varnost in tehnologija, d.o.o., Streliška ulica 3, 1000 Ljubljana, Slovenia

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


Paper presents the practical application of the Extended Best Estimate Plus Uncertainty (E-BEPU) methodology developed in the framework of the H2020 NARSIS project. The approach is risk-informed combined deterministic and probabilistic methodology dedicated to design verification. It assumes application of the best-estimate computer code, realistic input data for initial and boundary conditions with uncertainties and plant systems availability based on probabilistic safety analysis. The generic large Generation III Pressurized Water Reactor design defined in the NARSIS Project was applied, and the LB-LOCA postulated initiating event was studied as an example of design basis type event. The study is the first practical application of the proposed EBEPU methodology.

08.09.2020 15:40 Poster Session

Severe accidents - 1803

Assessment of Krško NPP emergency exercise and determination of source term

Benja Režonja, Tomaž Nemec, Tomi Živko

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


IAEA is preparing a new source term and accident scenario database. This database will contain pre-calculated source terms for different accident scenarios from National Competent Authorities. For this purpose, the SNSA has prepared calculations of source term for a hypothetical severe accident at the Krško NPP. The severe accident scenario was based on the emergency exercise NEK2019-2, conducted on 21st November 2019. The task was to prepare source term file in predefined format. Using generic IAEA TECDOC-955 methodology to assess the accident conditions, the source term was calculated for two different stages of the scenario.
The Accident assessment results that the SNSA prepared as an input to the IAEA database consists of:
• Initial plant state,
• Assumptions made on safety equipment operability throughout the transient,
• Core degradation kinetics,
• Evolution of specific plant parameters versus time.
To monitor the simulated plant parameters of the exercise in the plant, the SNSA employed the Emergency Response Data System (ERDS) connected to the plant simulator. The ERDS parameters are monitored in real time and include operation parameters of normal operating systems and safety systems, conditions of barriers, radiation and weather monitoring systems, warnings, alarms and status of critical safety functions.
To respond to an emergency at the Krško NPP, the SNSA has established an Emergency Response Center which is composed by Emergency director, Nuclear accident analysis group (source term evaluation), Dose assessment group (radioactive releases to environment, evaluation of doses to population) and Group of communicators (EMERCON, ECURIE, public). The Accident analysis group aim is to assess the plant conditions during an accident, to calculates a conservative source term for actual plant conditions and based on prognosis of a possible scenario development also a prediction of worst-case source term. The source term calculations are used as an input to the RODOS modelling that is performed by the Dose assessment group of the SNSA. The activity and release path of these predicted radioactive releases can be important as a basis for decision making on off-site protective measures in case of a release into environment.
The paper will present how the source term and accident scenario for the IAEA database were prepared. Two different radioactive releases were considered. The first one was a containment leakage through a penetration that was open for 30 minutes and thus released the isotopes unfiltered directly to the environment. The second release considered was the design leakage of the containment due to penetrations leakage within the limit defined in the plant’s Technical specifications.
The report was prepared using a thorough analysis of the ERDS data, results of assessment of exercise by the SNSA Accident analysis group and a predetermined exercise scenario supplied by the Krško NPP exercise planners. There were some challenges in preparing contributions for the IAEA database due to some unavailability of data and limited composition of source term. We can conclude that SNSA work in fulfilling the IAEA requirements despite all the limitations was still a success.

08.09.2020 15:40 Poster Session

Severe accidents - 1804

Simulation of LIVE2D Experiment on Reactor Core Melt in Reactor Pressure Vessel Lower Plenum

Blaž Kamenik1, Xiaoyang Gaus-Liu2, Jure Marn3, Ivo Kljenak4

1University of Maribor, Slomškov trg 15, 2000 MARIBOR, Slovenia

2Forschungszentrum Karlsruhe, Institute for nuclear and energy technologies (IKET) Hermann-von-Helmholtz-Platz-, Hermann-von-Heimholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

3Faculty of Mechanical Engineering, Aškerčeva 6, 1000 Ljubljana, Slovenia

4Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 LJUBLJANA, Slovenia


During a severe accident in a light-water reactor nuclear power plant, the reactor core melt could accumulate in the reactor pressure vessel (RPV) lower plenum. An accident management strategy in such cases consists in external cooling of the RPV that should significantly contribute to the removal of the residual heat from the melt. Both experimental and theoretical research is being carried out to understand and predict the melt behaviour, with the purpose of devising suitable mitigation strategies in case of accident. Essentially, the melt should be cooled down before the RPV fails and the melt flows into the reactor cavity. It is thus important to predict the heat flux from the core melt to the vessel walls.

Experiments with simulant materials, performed at the Karlsruhe Institute of Technology (Germany) in the LIVE2D facility, provide insights into the melt internal behaviour. The LIVE2D vessel has a semi-circular shape, with the width an order of magnitude lower than the diameter. In the vessel lower part, a salt melt represents the melt oxide layer, whereas a horizontal layer of oil on top of it represents the melt metallic layer. The salt melt may be heated with different heat flows that represent the decay heat. A solid crust may sometimes intermittently form and melt at the cooler boundaries (vessel semi-circular wall and upper surface).

Theoretical simulations of such experiments enable the validation of models to be used in actual plants, as well as support the analysis of experimental results. Two-dimensional simulations of some steady states or quasi-steady states during a LIVE2D experiment with the Computational Fluid Dynamics code ANSYS Fluent were performed. The simulated local temperatures are first compared to measurements. Then, the heat flow distributions over different parts of the semi-circular wall are analysed. Finally, the corresponding simulated flow patterns, which were not observed experimentally, are presented to provide new insights into the possible behaviour of the core melt.

08.09.2020 15:40 Poster Session

Severe accidents - 1805

Scaling Down of Experiment on Containment Atmosphere Mixing

Kristina Jovanovska1, Ivo Kljenak2

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


During a severe accident in a light-water reactor nuclear power plant, hydrogen may be generated due to oxidation of reactor core components and flow into the containment. If the local hydrogen concentration exceeds the flammability limit, hydrogen combustion could occur, jeopardising the containment integrity. The mechanisms leading to homogeneous or non-homogeneous hydrogen distribution in the containment, including the mixing of the containment atmosphere, are being investigated both experimentally and theoretically. These investigations should enable a better understanding of the underlying phenomena, which should be useful for devising suitable mitigation strategies.

Dealing with experimental results obtained in vessels of a volume of the order of 100 m3 has several open issues. One of them is the extrapolation of the findings to real containments with a volume of the order of a few tens of thousands m3. This is part of the wider topic of scaling of fluid mechanics phenomena, or somewhat narrower topic of scaling of fluid mixing in large volumes.

To investigate this topic the following approach was used. We chose two experiments performed in the PANDA experimental facility that is located at the Paul Scherrer Institute in Switzerland. During these experiments, the breaking up of a horizontal helium layer as a substitute for hydrogen, in a vessel with 90 m3 volume, was caused by natural convection induced by an electric heater. Scaled-down processes in models of ten times smaller volume were simulated with the Computational Fluid Dynamics software ANSYS CFX 18. In the experiments, the heater was placed at two different elevations, which resulted in different times necessary to break up the helium layer. The ratios of these times in the experiments and the simulations were compared to investigate the consistency of the phenomena in the scaling down. This was done considering the heating of the gas near the heater, the establishment of the natural convection and the breaking up of the helium layer. Some parameters were varied to investigate their influence on the consistency of the scaling, which may help to prescribe adequate experimental conditions in the future in order to get results that could be scaled-up to real containments.

08.09.2020 15:40 Poster Session

Severe accidents - 1806

Quenching of melt fine fragments during energetic fuel-sodium interactions

Mitja Uršič, Matjaž Leskovar

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


In the frame of safety studies for the sodium cooled fast reactors utilizing oxide fuels, the risk for the environment in the case of a vapour explosion must be estimated. A vapour explosion is an energetic event that can occur during the core melt accident when the rapid and intense heat transfer follows the interaction between the molten material and the coolant.
For the vapour explosion modelling, the pressurization modelling is important. The purpose of the pressurization modelling is to consider the heat transfer between the melt fine fragments and the surrounding coolant. Due to significant differences between the sodium and water thermal conductivity, the quenching of melt fine fragments might be considered almost instantaneously for sodium when compared to water. However, for both coolants the quenching rate of the melt fine fragments might be importantly affected by the slower heat transfer inside the melt.
Due to the effect of the melt thermal conduction, the temperature profile modelling inside the fine fragments might be considered important for the fuel-coolant interaction codes. But, in the explosion applications of current fuel-coolant interaction codes this temperature profile modelling is not considered. Thus, the objective of the paper is to discuss how the heat transfer limitation of the melt conductivity should be indirectly considered in the explosion phase simulations.

08.09.2020 15:40 Poster Session

Severe accidents - 1807

An Analysis of Combustion Regimes for Hydrogen/CO/Air Mixtures in Different Geometries

Mike Kuznetsov1, Andreas Friedrich2, Anke Veser2, Gottfried Necker2, Wolfgang Breitung1

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


During a MCCI accident in a reactor of NPP, within the subsequent ex-vessel phase, several hundred kg of H2 and CO can be produced. Then, a mixture of hydrogen and carbon monoxide in air can be formed as a stratified layer at the top of the reactor building. In presence of an ignition source, different flame propagation regimes for such a mixture may occur. The severity and danger of such combustion process in terms of maximum combustion pressure and temperature will depend on the geometry of the containment, scale, and composition of the combustible mixture.
A number of experiments with H2-CO mixtures at different ratios H2:CO in different geometries were performed at the HYKA test site of the KIT with respect to safety management of NPP. A series of experiments with H2:CO mixtures in air was performed in a spherical explosion chamber with a volume of 8.2 liters in order to evaluate the lower flammability limits and the laminar burning velocity. Such experimental data allow to theoretically predict the criteria for flame propagation regimes based on critical expansion ratio, ?*. The detonability limits were evaluated using detonation cell size according to 7? criterion.
A series of middle scale experiments in a tube and in a layer geometries was performed to experimentally confirm the theoretical evaluations of flame acceleration conditions based on critical expansion ratio. First, the experiments were performed in a 7.2-m long tube of 100 mm id with obstacles (blockage ratio was 30%) and second, in a horizontal semi-confined layer with dimensions of 9x3x0.6 m with/without obstacles opened from below. The ratio of H2:CO in test mixtures with air was varied as 3:1, 1:1, 1:3 to assess the influence of CO on flame propagation regimes. The experimental data can be used as benchmark experiments for numerical code validation.

08.09.2020 15:40 Poster Session

Severe accidents - 1808

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-20 with BWR bundle

Alexander D. Vasiliev

Nuclear Safety Institute of Russian Academy of Sciences , 52, B. Tulskaya, 115191 Moscow, Russian Federation


The QUENCH-20 test conditions simulated a severe LOCA (Loss of Coolant Accident) NPP (nuclear power plant) sequence in which the overheated up to 2000K core would be reflooded from the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. This test was performed in the frame of the SAFEST project in the cooperation with Swedish Radiation Safety Authority, Westinghouse Sweden, GRS and KTH and supported by the KIT program NUSAFE.
The QUENCH-20 experiment included the following phases: first heat-up phase, pre-oxidation phase, final heat-up phase and bottom water flooding phase. The test was successfully conducted at the KIT, Karlsruhe, Germany, in October 9, 2019.
The QUENCH facility is designed for studies of the nuclear reactor fuel assemblies behaviour under conditions simulating design basis, beyond design basis and severe accidents. The QUENCH-20 bundle simulates the BWR Nordic-type geometry with 24 electrically heated rods and B4C control blade.
The important feature of the QUENCH-20 test was the release of carbon monoxide CO, carbon dioxide CO2 as well as methane CH4 in the course of B4C oxidation. As a result of the test, the CO, CO2, and CH4 integral productions in QUENCH-20 test were 12.6, 9.7 and 0.4 g respectively. Totally 57.4 g of hydrogen H2 were released including 10 g from oxidation of B4C.
The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modelling code SOCRAT/V3 has been used for the calculation of QUENCH-16 experiment. The considerable challenge encountered in modelling was the complicated geometry of the bundle, which included, for example, two big rectangular absorber blades positioned at two connecting sides of the bundle square. In particular, this geometry is difficult for adequate modelling of radiative heat exchange in the bundle. However, the SOCRAT code has some flexibility to model radiative heat fluxes in rather complicated geometry. The calculated results are in a reasonable agreement with experimental data which justifies the adequacy of modelling capabilities of SOCRAT code for application to such a complicated test as QUENCH-20.

08.09.2020 15:40 Poster Session

Severe accidents - 1814

Development of the new ASYST Integral Analysis BEPU Severe Accident Analysis Code – Designing to Minimize the Influence of User Effects

Chris Allison

Innovative Systems Software, LLC, 1284 South Woodruff, Idaho Falls, Idaho 83404, USA


C.M. Allison1, J.K. Hohorst1, A. Ezzidi2, S. Jiang3, Z. Fu3, M. Perez-Ferragut4, H. Sánchez Mora5

1Innovative Systems Software (ISS)
2 Nuclear Power Engineering Consulting (NUPEC),
3Future RHYS,
4Energy Software Services (ENSO)
5Instituto Politécnico Nacional ESFM


ASYST (Adaptive SYStem Thermal-hydraulics) is a new code being developed by members of the ADTP (ASYST Development and Training Program) that combines the detailed severe accident analysis capabilities of SCDAPSIM and SAMPSON. The thermal hydraulic module, ASYST-THA, replaces the original US NRC-developed RELAP5 code used in RELAP/SCDAPSIM/MOD3.x and THA used in SAMPSON with new system level modeling options that include multi-dimensional, multi fluid models. The ASYST reactor-specific modeling options include modules describing the behavior of (a) the core/fuel assembly structures, (b) late phase debris/melt relocation, (c) the containment including melt spreading and molten core-concrete interactions, and (d) fission product release and transport.

The motivation for the development of ASYST arises from three related factors: (a) the creation of a unique detailed best estimate integral accident code that will offer a good balance between speed and accuracy that is not currently available for current and advanced reactor designs, (b) the elimination of the large impact of user effects that are observed because of the current necessity to use simplified parametric integral codes like MAAP and MELCOR for events as complex as those occurring in the Fukushima Daiichi accident, and (c) the creation of a framework for the longer term development of advanced modeling concepts that will gradually replace the current generation of detailed system codes like RELAP/SCDAPSIM and SAMPSON.
This paper will focus on the second objective of this project, the elimination of the large impact of user effects that was observed in the analysis of the Fukushima Daiichi accident. As noted in the full paper, the negative influence of user effects in system thermal hydraulic and severe accident codes has been widely demonstrated and discussed over the past 50 years. For example, in a variety of “blind” international standard problem (ISP) exercises performed over this time, large variations in predicted behavior between different codes, as well as different users of the same codes, have been observed. Although historically many of these differences have been attributed to the variation in modeling approaches, the adequacy of user guidelines, and the experience of the user, the experience from Fukushima clearly demonstrates that these issues have not been resolved even for mature and widely used codes. As discussed in this paper, ADTP will address this issue, through a multi-faceted approach including (a) the use of detailed and well proven models and correlations in ASYST that have been validated over a wide range of conditions, (b) improvement of the input to eliminate unnecessary model parameter options that have been retained primarily for historical reasons, (c) expand the capabilities of the integrated uncertainty analysis originally developed for SCDAPSIM to include default uncertainty distributions to address modeling uncertainties that are been defined through decades of model validation activities, (d) expand the capabilities of the advanced new desktop simulator GUI environments like GRAPE, Adv3Dgui, and RHYS to help eliminate common input errors and better understand calculated behavior for complex systems, and (e) improve the user guidelines and training to include user certification and expansion of reference input models and training materials to include the analysis of a wider range of accidents and NPP designs as well as the analysis of historic and ongoing integral experiments.

08.09.2020 10:30 Reactor physics

Reactor physics - 212

Noise analysis techniques of in-core modulation experiments of the European CORTEX project

Klemen Ambrožič, Vincent Lamirand

Swiss Federal Institute of Technology (EPFL), Station 3, Lausanne, Switzerland


Neutron flux fluctuations result from oscillations of various structures inside the nuclear reactor, which may be a consequence of boiling, coolant flow or seismic activity. These oscillations can be measured and localized using various neutron detectors. Study of these effects is the primary concern of the H2020-Cortex project, with the aim for development and validation of computational tools, which can be applied to commercial plants for core monitoring.
In order to provide experimental data for code validation, neutron flux fluctuation measurements have been performed in the EPFL Crocus reactor and the TUD AKR2 reactor, utilizing oscillating fuel assemblies and oscillating absorbers with adjustable oscillation shape and frequency. Multiple neutron detectors were used at various strategic positions in order to capture propagation dependence of the induced oscillation.
Power spectrum density and phase shift with respect to a reference detector and to the induced oscillation have been selected as measured and computed quantities of interest. In this paper we discuss the analysis steps taken in order to provide these quantities from the experimental data.
Initial step of the data analysis calls for data normalization and removal of low frequency components, which can be performed utilizing a moving averaging high pass filter with appropriate window length.
The measured reactor noise frequency coupling between individual detectors and oscillation itself is performed by coherence estimate in the frequency domain. Frequency power spectral density (PSD) and phase shift estimates are used for estimates of the oscillation power and distance from the oscillation and other detectors. PSD estimate ratios to a single reference detector normalized signal have provided valuable insight in the reactor behaviour and measurement responses to different oscillations frequencies and locations.
In order to estimate the measurement uncertainty, bootstrapping with replacement is employed using either a fixed or a randomly variable sample size. In addition to potential uncertainties, bootstrapping identifies potential biases as well.

08.09.2020 10:50 Reactor physics

Reactor physics - 213

Multigroup and continous energy MCNP analyses of the VVER-440 and LR0 reactors

Stefan Cerba, Jakub Lüley, Filip Osuský, Branislav Vrban, Vladimir Nečas

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


Following on the activities of the STU research team regarding the development of cross-section libraries for fast and thermal reactors, the V2020T version of the SBJ XS library has been developed. It comes with various improvements, such as being capable of preparing continuous energy and multigroup XS libraries for Monte Carlo calculations, as well as multigroup XS libraries for deterministic calculations, using the same processing options and evaluated data. However, the most significant improvement lies in the implementation of the latest CAB S(?,ß) scattering law model for hydrogen in light water. This paper presents the update of the SBJ XS processing scheme, which uses the latest NJOY21 nuclear data processing system embedded in an automated C++ utility. The carefully prepared set of multigroup and continuous energy cross-section libraries are used for criticality calculations of the VVER-440 power reactor and LR0 zero power research reactor. In case of VVER-440 heterogenous 2D core model is used and the estimated integral and differential parameters of the core are compared with those from the VVER-440 fuel assembly power distribution benchmark. For the LR0 zero power reactor detailed 3D core model is used and the emphasis is placed on estimating the influence of different flux weighting options on integral parameters and whether the used set of background cross sections is appropriate for this type of rectors. The calculations are performed using continuous energy and multigroup SBJ cross-section libraries in 238 group energy structure, processed based on ENDF/B-VII.1 evaluated data and the latest CAB S(?,ß) scattering law model.

08.09.2020 11:10 Reactor physics

Reactor physics - 211

Evaluation of angular burnup in the JSI TRIGA fuel element

Anze Pungercic, Luka Snoj

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


In the scope of reactor physics, determination of nuclear fuel burnup is of great importance from the standpoint of radiation fields characterization, safety and safeguards. Due to the difficulty of performing fuel burnup measurements, reactor calculations are used to determine the isotopic composition at different time-steps in reactor operation. With the recent increase in computer power, 3D Monte Carlo burnup calculations are possible, in which nuclear fuel is divided into multiple depletion zones to take into account the difference in local neutron flux and spectrum, which enables us to study in detail spatial burnup effects.
Serpent 2 Monte Carlo transport and burnup code was used to calculate burnup of the JSI TRIGA fuel elements. In the scope of the whole detailed fuel burnup analysis, angular spatial burnup effects were studied by dividing the fuel element by angle into 20 equal (18°) depletion zones. The interest was mainly in studying how the fuel element surrounding effects its angular burnup distribution. We have studied one fuel element in the core periphery and observed that the burnup is higher on the side of fuel element facing towards empty outermost ring filled with water. In order to analyse this effect, angular neutron spectrum changes were calculated using Monte Carlo calculations. It was observed that the thermal peak in region facing the higher amount of water is relatively higher, compared to other regions, which directly results in higher burnup and thus proving that fuel element surroundings notably effect the fuel burnup.
As the TRIGA core has annular configuration which is not periodic, the amount of water surrounding each fuel element depends on the FE location. Moreover the amount of water around fuel element also varies significantly with angle. In order to evaluate this effects we have used simple ray-tracing algorithm to determine the amount of water and compare the variations to fuel burnup. Relatively good agreement was observed. In the paper results for multiple fuel elements will be presented and connection between amount of water and fuel burnup evaluated, which could be used in the future for improvement of reduced order modelling methods such as conventional deterministic calculations or hybrid calculations using the fission matrix method.

08.09.2020 11:30 Reactor physics

Reactor physics - 214

Optimization of BEAVRS PWR loading pattern using a novel genetic algorithm based on population variance control

Wojciech Kubiński1, Piotr Darnowski1, Kamil Chęć2

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


The report presents a novel genetic algorithm (GA), and numerical tools developed to improve fuel cycle performance and in-core fuel management process. The primary purpose was to develop GA dedicated to solve core loading pattern (LP) problem with predefined core operation constraints. Particular focus was put on maximizing the length of the fuel cycle, but presented solutions allow to limit or control the magnitude of excess reactivity, population and type of fuel assemblies, fissile material mass and inventory of burnable absorbers. GA was implemented in a new computational framework written in Python3 and coupled with PARCS3.2 core simulator. The Pressurized Water Reactor (PWR) based on the MIT BEAVRS Benchmark was used as the demonstration case, and developed tools were applied to improve the first fuel cycle. Several test simulations were performed, including symmetry impact, population, mutation and other effects. New variance control method was proposed and tested. The obtained results for BEAVRS first fuel cycle show that new algorithms allow designing efficient fuel loading pattern. The most successful test case increased the length of the first BEAVRS fuel cycle by 94 days (or 28%) using almost the same mass of fissile materials, burnable absorbers and limiting excess reactivity to preserve the reactivity margin for control banks.

08.09.2020 11:50 Reactor physics

Reactor physics - 215

Criticality analysis of proposed TRIGA spent fuel pit design for various hypothetical events

Sefa Bektaş

Istanbul Technical University, Maslak, 34467 Sariyer/Istanbul, Turkey


Spent fuel pits are designed to store spent nuclear fuel elements temporarily after removal from the reactor core. They must have a carefully designed basket that indicates the position of the fuel elements in the pit to prevent criticality. This article presents the criticality analysis of different basket designs -filled with three different types of TRIGA fuels individually- for a proposed TRIGA spent fuel pit based on the Monte Carlo method using MCNP6.2 code because there is a direct link between libraries and the accuracy of methods employed to examine operations with fissionable materials. Two different basket designs having circular and hexagonal arrangements to store the different number of fuel elements in the pit were considered for criticality calculations. In addition, the effect of water density, the water level in the pit, and the replacement of light water with heavy water on the criticality was investigated to finalize the basket designs. The 3-D model of the baskets was generated for simulations. The results showed that the designs result in criticality in the pit under normal operating conditions which in compliance with the regulations. Even for the hypothetical events such as changing of water density and decrease in the water level in the pit, the effective neutron multiplication factors, keff remained below 1.0. Only, the replacement of light water with heavy water can cause criticality in the pit. It is verified that the design criteria are fulfilled and subcriticality is maintained for proposed TRIGA spent fuel pit in terms of criticality.

08.09.2020 12:10 Reactor physics

Reactor physics - 216

Validating the Serpent-Ants calculation chain using BEAVRS fresh core HZP data

Ville Valtavirta

VTT Technical Research Centre of Finland Ltd., P.O. Box 1000, FI-02044 VTT, Finland


VTT Technical Research Centre of Finland Ltd (VTT) is in the process of renewing its reactor analysis tools. The nodal neutronics code Ants has been developed since 2017 as a reduced order solver for the Serpent based two step reactor physics calculation chain. The in-house development of both solvers of the calculation chain allows the use of advanced group constant generation approaches for increased accuracy at the reduced order stage.

The rectangular nodal diffusion solution of Ants has been previously validated against numerical benchmarks in [1] and the rectangular pin power reconstruction methodology has been validated against Serpent using 2D configurations in [2]. In this paper, we will use data from the BEAVRS benchmark [3] for fresh core HZP conditions to validate Ants: The Serpent-Ants and 3D Serpent reference solutions for critical boron content (CBC) at different control rod insertion patterns will be compared to the measured data reported in the BEAVRS benchmark. Furthermore, the reactivity worths of each control rod bank predicted by Serpent-Ants and Serpent will be compared to the measured data. Finally, the assembly power and pin power distributions predicted by Serpent-Ants will be compared to the reference 3D full core Serpent solution.

In order to estimate the effects of some advanced group constant generation approaches, the following variations in the process will be considered:

*The use of the Monte Carlo based cumulative migration method (CMM) approach [4] for evaluating the diffusion coefficients of the fuel assemblies instead of the leakage corrected out-scatter diffusion coefficients.
*The application of fundamental mode (FM) leakage correction in the generation of the fuel assembly group constants.
*The correction of the discontinuity factors at the core-reflector interface based on combination of data from a 2D full core Serpent simulation, infinite lattice fuel assembly homogenization and single node calculations with the nodal solver as described in [5].
*The use of subnodalization, i.e. the division of fuel assemblies into 2x2 nodes with either:
i) The full assembly group constants being used for each quadrant or
ii) Group constants generated separately for each quadrant.

We will present the effects of these variations on the predicted CBCs, control rod bank worths as well as assembly, node and pin powers.

[1] V. Sahlberg and A. Rintala, “Development and first results of a new rectangular nodal diffusion solver of Ants”, In Proc. PHYSOR 2018, Cancun, Mexico, April 22-26, 2018.
[2] A. Rintala and V. Sahlberg, "Pin Power Reconstruction Method for Rectangular Geometry in Nodal Neutronics Program Ants", In Proc. NENE 2019, Portoroz, Slovenia, September 9-12, 2019.
[3] N. Horelik, B. Herman, B. Forget, and K. Smith. “Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS), v1.0.1.” In Proc. M&C 2013. Sun Valley, ID (2013).
[4] Z. Liu, K. Smith, B. Forget, and J. Ortensi. “Cumulative migration method for computing rigorous diffusion coefficients and transport cross sections from Monte Carlo.” Annals of Nuclear Energy, volume 112, pp. 507 – 516 (2018).
[5] K. S. Smith. “Nodal diffusion methods and lattice physics data in LWR analyses: Understanding numerous subtle details.” Progress in Nuclear Energy, volume 101, pp. 360 – 369 (2017). Special Issue on the Physics of Reactors International Conference PHYSOR 2016: Unifying Theory and Experiments in the 21st Century.

08.09.2020 09:10 Research reactors

Research reactors - 307

Impurity identification in lead-bismuth eutectic nuclear coolant by thermal cycling

Kristof Gladinez1, Jun Lim2, Kris Rosseel2, Alexander Aerts2

1SCK.CEN, Av. Herrmann Debrouxlaan 40, 1160 Brussels, Belgium

2CEN/Serma - Lepp, BAT.470, 91191 GIF-SUR-YVETTE, France


MYRRHA, a Multi-purpose hYbrid Research Reactor for High-tech Applications, is under design at the Belgian Nuclear Research Centre (SCK CEN). MYRRHA is a new generation sub-critical nuclear reactor driven by a proton accelerator. A proton beam accelerated up to 600 MeV impacts inside the nuclear reactor core on heavy liquid metal atoms, lead and bismuth. This heavy metal spallation target generates neutrons to maintain the sub-critical fission reaction in the core. Due to this configuration, the MYRRHA reactor is a so-called Accelerator Driven System (ADS) able to transmute long-lived minor actinides. The primary coolant of the MYRRHA reactor is chosen to be an eutectic mixture of lead and bismuth (LBE) and is therefore suitable as the spallation target as well .
It is well-known that accurate control of the primary coolant cleanliness is of prime important in nuclear reactors. Light-water reactors are known to be prone to CRUD formation. Although chemical reactions in LBE are vastly different, solids formation and subsequent deposition is considered to be extremely important. First of all, an accurate control of the dissolved oxygen concentration in LBE is considered necessary to limit corrosion of structural steel and to mitigate coolant oxidation or fouling by lead oxide. Additionally, stainless steel elements such as iron (Fe), Nickel (Ni) and Chromium (Cr) are released in the liquid metal coolant due to corrosion. High concentration of these elements might lead to precipitation of solid compounds in cold spots in a nuclear system. To avoid such processes, the control of dissolved impurities is necessary. A first start of dissolved impurity control is based on establishing accurate knowledge on the chemical interactions of these elements in the liquid metal.
Successive cooling and re-heating of LBE from 450°C to 200°C at a fixed total oxygen content is performed to evaluate the formation of metal oxides in LBE. During these cycles, the dissolved oxygen content is measured using potentiometric oxygen sensors. A decrease in the measured dissolved oxygen during cooling of LBE indicates the formation of (metallic) oxides in the liquid metal. By performing such thermal cycles at different oxygen and impurity content, oxide formation can be quantified. The thermal cycling experiments performed at various oxygen content in LBE have indicated possible iron and iron-nickel oxide formation in temperature ranges of interest for MYRRHA. The experimentally obtained results are compared to thermo-chemical predictions of oxide formation in LBE. The results agree rather well for the formation of magnetite, hematite and iron-nickel oxide in LBE.

08.09.2020 09:30 Research reactors

Research reactors - 308

Conventional fast neutron flux measurements in the radial piercing channel D of the TRIGA Mark II reactor, Pavia

Marco Di Luzio

Istituto Nazionale di Ricerca Metrologica, Viale Taramelli 12, Pavia, 27100, Italy


The TRIGA Mark II nuclear reactor operated by the Laboratory of Applied Nuclear Energy (LENA) of the University of Pavia is a 250 kW light water moderated facility aimed for training, general purpose research and isotope production [1].
One of the research fields conducted at this reactor is focused on the investigation of nuclear reactions induced by fast neutrons. The accessibility to a fast neutron beam allows a broad variety of applications, ranging from the determination of fast neutron cross section data, study of burnup and transmutation in fuel elements and effects of radiation damage in various materials; all these applications might be, in turn, beneficial for what concerns research and development of the upcoming IV generation of fast nuclear reactors.
In order to make available a fast neutron beam at the TRIGA reactor, the realization of a new neutron irradiation facility is planned by modifying the so-called channel D, a pre-existing radial piercing channel without reflector material. The channel D will be adapted by introducing filters to remove the neutron thermal flux component and reduce the gamma background, and a beam catcher to assure operator’s safety. Characteristics and dimensions of filtering and shielding materials will be modulated according to the neutron flux spectra simulated by means of Monte Carlo Neutron Particle (MCNP) software. The quality of the simulated data will be assured by validating the software code using experimental data collected in selected positions along the channel.
In this study, preliminary measurements of conventional fast neutron flux in four positions of the actual (unmodified) configuration of channel D are reported. The adopted technique, described in [2], consists of activations and ?-countings of monitor elements. The resulting fast neutron flux relies on a conventional value of the monitor fission-neutron averaged cross section, ?. Ni solutions, obtained from a Ni solid standard dissolved in nitric acid, were selected as monitors to exploit the 58Ni(n,p)58Co threshold reaction; equal volume samples of four solutions prepared with increasing Ni concentration were placed at 45 cm, 75 cm, 125 cm and 195 cm from the vertical axis of reactor core. The neutron exposure lasted 90 minutes at 10 kW power; the absence, during the measurement, of most part of the shielding prevented the achievement of the operational 250 kW power. Gamma spectra including the 58Co 810.8 keV ?-peak emission were acquired with a calibrated HyperPure Ge (HPGe) detector by placing the irradiated samples in contact with the HPGe end-cap.
The conventional fast flux, ?f, for each measurement position was obtained using eq. (10) of [3]; we also calculated the reaction rate per target nucleus in Ni monitors, R, by multiplying the ?f times ?. The R value is no longer conventional as it is independent from assumptions concerning the fast neutron flux shape. Since the R values can be evaluated by MCNP code, they are used to assess the quality of simulated data.
The ?f and R results, scaled at 250 kW reactor power, ranged from 1.33(4) x 10^11 cm^-2 s^-1 and 1.48(3) x 10^-14 s^-1, respectively (at 45 cm), to 1.32(4) x 10^9 cm^-2 s^-1 and 1.46(3) x 10^-16 s^-1, respectively (at 195 cm). Values in parenthesis indicate the standard uncertainty. The two orders of magnitude decrease in flux and reaction rate along the 150 cm horizontal distance was in agreement with previous knowledge of the facility.
To sum up, the experimental data collected in this preliminary measurements offer a valuable independent reference to validate the code that will be used for modeling the structural modifications on channel D.

[1]Prata et al; Eur Phys J Plus, 2014; 129.
[2]De Corte; habilitation thesis, University of Gent, 1987.
[3]Di Luzio et al; Prog Nucl Energy, 2019; 113.

08.09.2020 09:50 Research reactors

Research reactors - 309

Sensitivity analysis of stainless steel reflector for VR-1 training reactor

Jan Frybort1, Pavel Suk2, Lubomir Sklenka1, Filip Fejt2, Lenka Frybortova2

1Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors, V. Holesovickach 2, 18000 Praha 8, Czech Republic


Pressurized water reactors are typically surrounded in the radial direction by neutron reflectors made from stainless steel and water. These reflectors decrease neutron leakage and provide protection of pressure vessel from fast neutrons damaging its integrity. Such a radial reflector influences multiplication factor of the core and distribution of neutron flux and fission power inside the core. All these effects can be analyzed by full-core simulations using macroscopic constants. Methodology for generation of the macroscopic constants for non-fuel regions will be tested for new stainless steel reflectors at the VR-1 reactor. Rods from SS 304l material will be used for construction of radial reflectors for the VR-1 reactor. They will be design to generate sufficient measurable response in selected core characteristics.
The study is focused on core power distribution and reactivity worth of absorbing rods in a VR-1 reactor core. The core typically consists of about 20 IRT-4M fuel assemblies and seven absorbing rods UR-70. Replacing water surrounding the core by several reflector assemblies containing stainless steel will influence leakage and distribution of neutrons inside the core. The current analysis deals with local effects and employs the sensitivity study to discover the nature of reflectors' impact on the reactor core. These effects were studied even for several past VR-1 reactor core configurations. All calculations were carried out in Serpent2 Monte-Carlo code with various evaluated libraries: ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 data.

08.09.2020 14:00 Thermal-hydraulics, computational fluid dynamics

Thermal-hydraulics, computational fluid dynamics - 713

Experimental investigation on debris bed quenching with additional non-condensable gas injection

Markus Petroff, Rudi Kulenovic, Joerg Starflinger

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


Severe accidents of light water reactors with a loss of coolant can result in overheating of the fuel rods and the loss of core integrity. In case of insufficient cooling, the reactor pressure vessel may fail and the discharged melt from the vessel will fall into the reactor cavity and form a structure of heat-generating particles of different shapes and sizes (debris) by fragmentation in the residual water. In the scenario of depleted residual water, the melt will interact with the concrete underneath generating gas at the bottom of the particle bed (MCCI), which will flow through the debris bed.
Experimental investigations have shown that non-condensable gases (NCG) can slow down the quenching process. Fundamentally, the impact of additional gas on the quenching process is numerically modelled in thermal hydraulic system codes like ATHLET, however there is still a need for experimental validation of respective models or verification of corresponding simulation results. Therefore, especially for the model validation of COCOMO-3D implemented in ATHLET, a specific extension to the existing experimental database is required.
The paper gives a comprehensive outline on the state-of-the-art of previous experimental research related to debris bed quenching with involved NCG, and describes in detail the new test facility FLOAT, which has been established at the IKE, University of Stuttgart, for specific experiments on the coolability of particle beds with additional NCG injection from the bottom. First experimental results of the quenching behaviour of a monodispersed particle bed at top-flooding cooling condition are presented and the effects of an additional NCG flow (air) from the bottom of the particle bed are discussed.

08.09.2020 14:20 Thermal-hydraulics, computational fluid dynamics

Thermal-hydraulics, computational fluid dynamics - 714

Considerations on Successful Similarity Theories Developed for Flow Stability and Heat Transfer with Fluids at Supercritical Pressure

Andrea Pucciarelli, Walter Ambrosini

Dipartimento d'Ingegneria Civile e Industriale - Universita di Pisa, Largo Lucio Lazzarino, 1, 56122 - Pisa (PI), Italy


Similarity theories are necessary whenever trying to grab the most important characteristics of complex phenomena on the basis of relevant dimensionless numbers, representing in a compact form the boundary conditions and the key features of a problem. Supercritical pressure fluids exhibit such a complexity in behaviour, owing to the dramatic changes in their properties across the pseudocritical temperature, where the fluid, though being single-phase under a physical respect, is actually characterised by liquid-like and gas-like regions, to be discriminated by the large differences observed in density and other thermodynamic and transport properties.
The interest in supercritical fluids for nuclear technology stems from the conception of a Generation IV type reactor cooled by supercritical water (SCWR), proposed a few decades ago as one of the six designs on which research and development should be concentrated. Though since the time of this first proposals the concept has undergone several developments, still challenges are present in its evolution to a mature technology, a situation being a common feature for all Generation IV reactor concepts. In the particular case of SCWRs, in addition to problems related to the coolant compatibility with structural materials, owing to the attempt to reach higher temperatures and then higher energy conversion efficiencies, the difficulties encountered concern the mentioned changes in fluid properties, making extremely difficult to develop engineering correlations and turbulence models capable to provide a satisfactory prediction of heat transfer conditions in nuclear reactor cores and components. As a matter of fact, the combination of heavy and light phases transforms the fluid into a sort of two-phase one, though in the absence of interfaces. Therefore, though the adoption of supercritical operating pressures rules out worries about reaching critical heat flux (CHF) conditions, other unwanted phenomena are involved, being heat transfer deterioration (HTD) and possible static and dynamic instabilities, in close similarity to what observed for boiling water reactors (BWRs).
Notwithstanding the interest for supercritical fluids dated since the 60s in the last century and their extended applications even in conventional thermal power stations, the use of supercritical water as a coolant for nuclear reactors requires a level of reliability in modelling and predictive tools that asks for further improvements. A basic step in this regard is the development of the mentioned engineering heat transfer correlations and, considering present up-to-date research and design tools, of turbulence models for CFD that represent in a correct way the fascinating phenomena exhibited by supercritical fluids. In the frame of these studies, setting up fluid-to-fluid similarity theories is of overwhelming importance to promote understanding and ease the design of experimental facilities: this, considering current literature, has been up to now the ground for repeated attempts repaid by little overall success.
This paper presents a synoptic overview of successful similarity theories developed by the authors and collaborators in the last two decades for flow stability and heat transfer, which only recently reached a full maturity with the demonstration of the capability to be applicable in the heat transfer frame to a wide variety of fluids. Meaningful quantitative results are illustrated and the new perspectives opened by their application are thoroughly discussed.

08.09.2020 14:40 Thermal-hydraulics, computational fluid dynamics

Thermal-hydraulics, computational fluid dynamics - 715

Two-fluid model simulations of isothermal stratified counter-current flow of air and water with interface compression and turbulence damping

Matej Tekavčič1, Richard Meller2, Fabian Schlegel2, Boštjan Končar1

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

2Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden, Germany


Stratified flows of water and steam can appear in the primary system of a pressurized water reactor during a hypothetical loss-of-coolant accident. Among others, important safety concerns during cold water injection of the emergency core cooling system include the pressurized thermal shock and the possible formation of a condensation induced water hammer. Both mechanisms could cause significant thermal and mechanical stresses on the components of the primary system. Thorough knowledge of turbulent heat and mass transfer processes near the interface is required for safety analyses of both phenomena.

Measurements of industrially relevant turbulent two-phase flows tend to be difficult; therefore computational fluid dynamics simulations represent an important additional analytical tool. The main objective of the present research and development is to advance the capabilities of current state-of-the-art modeling tools towards the simulations of two-phase flow phenomena under realistic reactor conditions. In the present paper, the focus is on turbulence modelling near the gas liquid interface in stratified flows.

An isothermal stratified counter-current flow of air and water in a rectangular channel is simulated. Computational domain and boundary conditions are based on the flow conditions in the test section of the WENKA experiment [1]. The validation case considers supercritical stratified flow with Froude number of 2.36 and Reynolds number 12000 for water and 27000 for air.

The two-fluid modeling approach with interface compression is used to resolve the interface between the two phases. A consistent momentum interpolation numerical scheme is applied, featuring the partial elimination algorithm to handle the strong interphase drag coupling at a resolved interface. The Unsteady Reynolds Averaged Navier-Stokes (URANS) approach is used to describe turbulent two-phase flow. Modelling of turbulence dissipation at the interface requires a special treatment that includes introduction of additional turbulence damping terms into the k-? Shear Stress Transport (SST) turbulence equations. Simulations, model and source code development are performed with the open source C++ library OpenFOAM.

Simulation results are validated with the measured profiles of volume fraction, velocity and turbulent kinetic energy at two streamwise positions in the test section of the WENKA experiment. Results of the mesh sensitivity study are presented. Furthermore, results of a parametric study reveal that an asymmetric damping approach with a lower coefficient on the liquid side of the interface can improve the prediction of turbulent kinetic energy profiles.

[1] Stäbler, T., Meyer, L., Schulenberg, T., & Laurien, E. (2006). Turbulence Structures in Horizontal Two-Phase Flows Under Counter-Current Conditions. Proceedings of FEDSM2006 (pp. 61–66). ASME.

08.09.2020 15:00 Thermal-hydraulics, computational fluid dynamics

Thermal-hydraulics, computational fluid dynamics - 716

Processing of Thermal-hydraulic Approximations in the GFR 2400 Fast Reactor Design

Filip Osuský, Branislav Vrban, Stefan Cerba, Jakub Lüley, Vladimir Nečas

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


The paper investigates the transient processes in the gas-cooled fast reactor referred as GFR 2400. The NESTLE code is used as coupled simulation tool and solves multigroup neutron diffusion equation by finite difference method that is internally coupled with thermal-hydraulic sub-channel code. The NESTLE code uses during the calculation process an effective fuel temperature which is defined as the uniform temperature for which the neutron multiplication properties are equal to the multiplication properties for a specific radial temperature profile. To correctly determine temperature distributions in the fuel, the in-house developed code referred as TEMPIN is used. The TEMPIN code solves steady state heat balance equation with flowing coolant in triangular lattice cell and the thermal-hydraulic properties of the fuel pin together with the coolant are temperature dependent. The heat source is defined by the TRITON sequence from SCALE package system and is normalised by the defined linear power in the fuel pin. Based on calculated temperature distributions by the TEMPIN code, the thermal-hydraulic approximations suitable for the NESTLE code are processed. The ability of the NESTLE code is to determine nodal averaged thermal-hydraulic physical parameters what is not sufficient for the licensing purposes. Therefore, next application of the TEMPIN code is to determine maximal fuel temperate within the fuel bundle. The results of the analyses are compared with the previous study which used the FLUENT code (from ANSYS code package system) for processing of thermal-hydraulic approximations. Changes in neutronic and thermal-hydraulic distributions are described and visualized in the paper.

08.09.2020 15:20 Thermal-hydraulics, computational fluid dynamics

Thermal-hydraulics, computational fluid dynamics - 717

Flow dependent turbulent Schmidt number in containment atmosphere mixing

Rok Krpan, Iztok Tiselj, Ivo Kljenak

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


During a severe accident, a hydrogen explosion could threaten the integrity of the nuclear power plant containment, which could lead into release of radioactive material into the environment. Various experiments are performed to simulate physical phenomena occurring in containment during severe accidents and results are used to validate Computational Fluid Dynamics (CFD) codes in order to simulate phenomena in actual power plants.
The experiments on containment mixing performed in MiniPANDA, PANDA and SPARC experimental facilities are considered in the present work. The interaction of vertical axisymmetric air or steam jet on a horizontal layer of helium-steam or helium-air mixture is simulated with open-source CFD code OpenFOAM. Turbulent Schmidt number, which is included in the transport equation of gas species, is defined as the ratio of turbulent transport of momentum to the turbulent transport of mass. In CFD calculations a constant value of turbulent Schmidt number is usually used. Typically, turbulent Schmidt number values used in calculations of injection of fluid into a reservoir containing stagnant fluid are specified with the comparison of calculation and experimental results, and are in the range of 0.7 to 0.9. However, a constant turbulent Schmidt number value does not always provide the best results. In some cases the results still differ substantially from the experimental results. The changes in turbulent Schmidt number usually occur within the shear flow layer, which is located at the boundary of the jet. Consequently, different values of turbulent Schmidt number can be prescribed to different regions of the flow (main jet, returning jet, quiescent environment), which can be specified according to the mean flow properties. With higher turbulent Schmidt number value in the jet region and lower in the quiescent environment the underestimation of turbulent diffusion of momentum in the regions with low velocity is compensated with increased turbulent mass diffusion and the discrepancies between simulation and experimental results are minimized.

08.09.2020 16:20 Safety analyses

Safety analyses - 805

RELAP5/MOD3.3 simulation of LOFT LP-FW-1 total loss of feedwater test

Andrej Prošek

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


After Fukushima-Daiichi in the Europe the design extension conditions (DEC) were introduced as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. Both WENRA guidance document for issue F and IAEA document provides the total loss of feed water (LOFW) as an example of DEC. The purpose of this study was to assess the latest RELAP5/MOD3.3 Patch 05 computer code for the simulation of the total loss of feedwater. The LP-FW-01 test performed in 1983 on the Loss of Fluid Test Facility (LOFT) has been used for simulation. The RELAP5/MOD3 steady state input deck available from literature has been adapted to RELAP5/MOD3.3 Patch 05, while transient input deck to simulate LP-FW-01 has been newly developed. The LP-FW-01 test represents a fault sequence in which a total loss of feedwater to the steam generator is followed by recovery by primary system feed-and-bleed. The coolant is simultaneously injected by the high pressure injection system (HPIS) and vented via the primary side power operated relief valve (PORV).
The LOFT facility was a 50 MWth two-loop pressurized water reactor (PWR). It was designed to study the thermo-hydraulic response of the system to a variety of simulated loss of coolant accident (LOCAs) scenarios. The LOFT facility conducted 38 experiments, before it was shutdown in 1985. In October 2006 it was completed its decontamination, decommissioning and demolition. The LP-FW-1 test measured data have been obtained through Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) data bank. The objectives of LP-FW-1 test were to provide data for code assessment, to allow assessment of the effectiveness of using PORVs and HPIS injection to remove reactor decay heat until residual heat removal (RHR) conditions are approached during a total LOFW and to provide information on transient characteristics to aid operators in the identification of, and recovery from, a LOFW transient. This can support DEC analyses for existing nuclear power plants on the need of DEC safety features for total LOWF. Namely, the control of DECs is expected to be achieved primarily by features implemented in the design (safety features for DECs) and not only by accident management measures that are using equipment designed for other purposes.
The simulation results for steady-state, short term response (0-300 s) and long term response (0-7000 s) will be presented in the paper. The results suggest that in the short term simulation of LP-FW-1 test the simulated results matches the major events very well. In the long term the simulation results suggest that besides code models also input modelling of steam flow may be important for correct predictions.

08.09.2020 16:40 Safety analyses

Safety analyses - 806

Analysis of IFA-650.4 LOCA Test for the Validation of Integrated code of MARS-KS and FRAPTRAN

Chang-Yong Jin1, Joo-Suk Lee2, Deog Yeon Oh1, Byung Gil Huh2

2Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea


Nuclear fuel at high burn-up condition can fail at lower temperatures than specified acceptance criteria of emergency core cooling system (ECCS). Therefore, the acceptance criteria of ECCS has been newly proposed to take into account the effect of high burn-up fuel in Korea. When the proposed acceptance criteria are adopted, safety analysis of loss of coolant accident (LOCA) as a design basis accident (DBA) of ECCS should be performed to properly predict the behavior of high burn-up fuel. For the prediction of the behavior of system with high burn-up fuel, the integrated code was developed as a regulatory audit calculation code in which system analysis code, MARS-KS1.4 was combined with S-FRAPTRAN module based on fuel performance code, FRAPTRAN-2.0. For the validation of the integrated code, Halden IFA LOCA tests have been used because fuel behaviors including fuel fragmentation, relocation and dispersal (FFRD) were shown due to the high burn-up condition of Halden IFA LOCA tests. In this paper, Halden IFA-650.4 LOCA test was calculated by the integrated code since Halden IFA-650.4 LOCA test with extremely high burn-up conditions showed large ballooning and burst opening so that the test is proper to validate fuel model in the integrated code including fuel relocation model.

07.09.2020 17:20 New reactor designs and small modular reactors

New reactor designs and small modular reactors - 408

TEPLATOR: nuclear district heating solution.

Radek Skoda1, Martin Lovecky2, Jiri Zavorka2, Anna Fortova3, Michal Zeman1, David Masata4, Frantisek Kolar1, Eva Vilimova3, Tomas Peltan3, Jan Skarohlid1, Ondrej Burian5, Jana Jirickova1

2University of West Bohemia, Univerzitni 8, 306 14 Plzeň, Czech Republic

3University of West Bohemia , Univerzitní 8, 31600 Pilsen, Czech Republic

4University of West Bohemia, Faculty of Electrical Engineering, Univerzitni 8, 301 00 Pilsen, Czech Republic


The innovative concept for district and process heat production is presented using already irradiated nuclear fuel from commercial light water power reactors which is not burnt up to its regulatory and design limits. The concept has been developed by a dozen of scientists from top Czech research organisations.

TEPLATOR is a critical assembly derived by the state of the art computational tools using better moderation, more optimal fuel lattice pitch, lower fuel temperature, lower coolant pressure for producing commercial heat with a cost of less than 4 EUR/GJ. Investment cost for building the TEPLATOR district heating station is below 30M EUR (for both using prices of 2019).

Different TEPLATOR variants are proposed; using either used BWR, PWR or VVER irradiated fuel assemblies (FAs). TEPLATOR can also be operated with fresh fuel, if the stockpile of irradiated FAs are exhausted.

TEPLATOR DEMO is a 50 MWt district heating plant using 55 FAs from VVER-440, producing 98 C hot water. TEPLATOR DEMO is coupled to a thermal storage system allowing shaving off morning and evening district heating peaks. TEPLATOR DEMO coolant is used at atmospheric pressure, the system has three loops, three main circulation pumps, three heat exchangers and heat generation is regulated by standard control mechanisms.

TEPLATOR variants using BWR and PWR square lattice fuel were also considered. The engineering constraints show potential for a higher output ( < 250 MWt) and/or higher temperatures ( < 200 C) as customers require.

The TEPLATOR solutions is especially suitable for countries that have thousands FAs stored either in interim storage casks or spent fuel pools. These FAs are now financial liability which, once used for heat production, can turn into a sizeable financial asset.

07.09.2020 17:40 New reactor designs and small modular reactors

New reactor designs and small modular reactors - 409

Current Status of Nuclear Power in the World Including Latest Developments on SMRs

Igor Pioro

University of Ontario, Faculty of Energy Systems and Nuclear Science, Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario L1H 7K4, Canada


It is well known that electrical-power generation plays the key role in advances in industry, agriculture, technology, and the standard of living. Also, strong power industry with diverse energy sources is very important for a country independence. In general, electrical energy can be mainly generated from: 1) non-renewable energy sources (75.5% of the total electricity generation) such as coal (38.3%), natural gas (23.1%), oil (3.7%), and nuclear (10.4%); and 2) renewable energy sources (24.5%) such as hydro, biomass, wind, geothermal, solar, and tidal power. Today, the main sources for worldwide electrical-energy generation are: 1) thermal power (61.4%) – primarily using coal and secondarily using natural gas; 2) “large” hydro-electric plants (16.6%); and 3) nuclear power (10.4%). The balance of the energy sources (11.6%) is from using oil, biomass, wind, geothermal and solar, and have visible impact just in some countries.
The paper presents a current status of electricity generation in the world, various sources of industrial electricity generation, and role of nuclear power with a comparison of nuclear-energy systems to other energy systems. This is viewed in the context of competitive global energy markets, emerging concerns on climate change, and meeting rowing world energy demands in a sustainable manner.
Unfortunately, within last years, electricity generation with nuclear power has decreased from 14% before the Fukushima NPP severe accident in March of 2011 to about 10%. Therefore, it is important not just to evaluate current status of nuclear-power industry, but make projections on near term (5–10 years) and long term (10–25 years and beyond) trends. During this timeframe, it is important to develop innovative nuclear technologies that would become more cost competitive than today’s NPPs and with other sources of energy.
SMRs are today’s a very “hot” topic in nuclear engineering worldwide. According to the IAEA ARIS data, there are about 55 SMRs designs / concepts, which can be classified as: 1) Water-cooled SMRs (land based) ? 19; 2) Water-cooled SMRs (marine based) ? 6; 3) High-temperature gas-cooled SMRs ? 10; 4) Molten-salt SMRs ? 9; 5) Fast-neutron-spectrum SMRs ? 10; and 6) Other SMRs ? 1.
France, Russia, UK, USA and other countries have great experience in successful development, manufacturing, and operation of submarines, icebreakers, and ships propulsion reactors. Therefore, many modern designs / concepts of SMRs are based on these achievements. Also, it should be mentioned that a number of SMRs concepts are based on the Generation IV nuclear-power-reactors concepts.
From all these 55 SMRs only two KLT-40S reactors (Generation-III) have been constructed, installed on a barge, and put into operation in port of Pevek, Russia in 2019. Also, Russia has proven design of Generation-III+ SMR – RITM-200M, which is already manufactured and will be put into an operation in the nearest future.
In general, specifics of SMRs are: 1) Relatively small installed capacities up to 300 MWel; 2) Long term operation between refuellings; and 3) Fuel enrichments up to 20%.

07.09.2020 18:00 New reactor designs and small modular reactors

New reactor designs and small modular reactors - 410

Development of novel safety features in small modular reactors

Salah Ud-Din Khan

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


Small modular reactors (SMRs) are considered to be safer reactor design due to inherent safety concepts and passive safety features. Therefore, in this paper, two novel safety concepts were developed for SMRs. These safety features include passive safety systems concepts. At first, steady-state conditions of SMR with a power output of 100 MWth were obtained by computational techniques. Secondly, two residual heat removal systems were design and integrated with this SMR and calculate various parameters and analyze the performance of the reactor under accidental conditions. Neutron kinetics and thermal hydraulic-calculations were solved to calculate various parameters of the reactor operation. Modeling of passive decay heat removal system was performed for both concepts. Two simulation models were developed that can be used to authenticate the safety assessment of this particular SMR. The developed optimized simulation model can be very helpful for governing bodies to assess the safety performance of any SMR.

07.09.2020 18:20 New reactor designs and small modular reactors

New reactor designs and small modular reactors - 411

Experimental Infrastructure in Support of Licensing and RDI Activities for LFR and ALFRED Implementation in Romania

Marin Constantin1, Ilie Turcu2, Daniela Diaconu3, Minodora Apostol3, Mirela Nitoi4, Daniela Gugiu5

1Institute for Nuclear Research Pitesti, Romania, Campului 1, 115400 Mioveni, Romania

2Institute for Nuclear Research Pitesti, Campului 1,P.O. Box 78, 115400 Pitesti-Mioveni, Romania

3Institute for Nuclear Research Pitesti, Campului Street 1, Mioveni, 115400, Romania

4RATEN ICN Institutul de Cercetari Nucleare Pitesti, Str. Campului nr.1, 115400 Mioveni, Romania


Lead Fast Reactors (LFR) is one of the six technologies recommended by GIF for Generation IV development, and one of the three supported by SNETP. ALFRED, the demonstrator of LFR technology, is aimed to verify the technological viability and the economics. It is planned to be built in Romania, and Mioveni nuclear platform was selected as the reference site. An international consortium (FALCON) was set-up to coordinate all preparatory activities.
A detailed analysis of the licensing process was performed and the experimental needs were identified. Therefore a list of the infrastructure gaps was produced considering the existing experimental facilities and their availability. A list of six new experimental infrastructures resulted: ATHENA, ChemLab, ELF, HELENA2, HandsON, Melting’Pot. In order to contribute both to the licensing purpose and training activities these experimental facilities will be built on Mioveni nuclear platform. A dedication Hub is planned to coordinate the experimental activities, including the open access approach. Beyond the horizon of the licensing process dedicated activities to find appropriate solutions for the open issues of LFR technology were anticipated to support the sustainability of the investment. The six infrastructures are in different phases of the preparation. The paper is focused on the process of the implementation presenting the stage of the preparation of the technical documentation, the application for structural funds, or for the contracting. The paper analyses the advantages of the approaches, the efforts for the investment, estimated impact, and the implementation risks.

08.09.2020 15:40 New reactor designs and small modular reactors

New reactor designs and small modular reactors - 404

Thermal energy storage for TEPLATOR: technology, utilisation and economics

Jan Škarohlíd1, Ondrej Burian2, Radek Skoda1, Michal Zeman1, Anna Fortova3

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


Generally, Energy storage is a very current topic nowadays as renewable sources of energy produces more cheap but unpredictable energy. Energy storages became more and more common not even for electricity but also for heat. Thanks, cheap energy and stronger and smarter control systems and consumption model’s energy storage is becoming not only as needed solution but also as a wanted solution even as energy storage for district heating.
TEPLATOR is a critical assembly using already irradiated nuclear fuel from commercial light water power reactors which is not burnt up to its regulatory and design limits. This innovative concept for district heating could benefit from having a decent heat energy storage for compensation of: 1) TEPLATOR power fluctuations, 2) Compensation and smoothing of the demand curve and 3) can serve as an emergency and safety heat sink.
Molten salt heat storage is a promising solution, which operates in a suitable temperature range, could absorb adequate amount of heat in reasonable material volume and with good operation dynamics providing quick response for charging and discharging demands.
In this paper we would like to point out and discuss benefits, possibilities and economics of Molten salt heat storage operating with TEPLATOR.

09.09.2020 11:10 Radiation and environmental protection

Radiation and environmental protection - 608

Remote radiation inspection of Jožef Stefan Institute Mark II TRIGA reactor using a mobile robotic platform

Ioannis Tsitsimpelis1, Andrew West2, Anže Jazbec3, Luka Snoj4, Barry Lennox5, Phillip A. Martin5, Malcolm Joyce1

1Lancaster University, Department of Engineering, Bailrigg, Lancaster, LA1 4YW, United Kingdom

3Jožef Stefan Institute, Reactor Infrastructure Centre, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

4Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

5School of Mechanical, Aerospace and Civil Engineering, The University of Manchester , Manchester M13 9PL, Manchester M13 9PL, United Kingdom


Research on the integration of robotic platforms with radiation detection capabilities has intensified in the past decade, partly due to the catastrophic events at the Fukushima Daiichi nuclear power plant, but also due to fundamental advances in autonomy and user-friendliness of operating robotic systems. The outputs of this global effort have the potential to provide solutions in a wide range of critical tasks, spanning from remote routine inspections, to enabling collection of information from inaccessible environments, and increasing the range of survey options in emergency situations. Eventual adoption of this technology will be rewarding for the nuclear industry, particularly with respect to the safety of radiation workers, the consistency and completeness of the information that operators can obtain, and the clarity with which operators can respond to enquiries from regulatory bodies. The research presented in this article is driven by this vision and contributes to the concept of remote radiation inspection. Within the framework of the TORONE project (Total Characterisation and Remote Observation of Nuclear Environments), an unmanned ground vehicle with Robot Operating System (ROS) enabled radiation detection instrumentation has been developed. By exploiting gamma radiation measurements, it has been possible to carry out in-situ pulse height spectroscopy, radiation imaging of targeted areas using a single detector-collimator arrangement and high-resolution 2D maps of radiation fields. The vehicle is equipped with state-of-the art LiDAR sensors, imaging cameras, and can be operated remotely via wireless link. This article provides an overview of the various subsystems and their integration, as well as the capabilities of the overall system, and it presents deployment results from the first mobile characterisation of a TRIGA reactor, at the Jožef Stefan Institute.

09.09.2020 11:30 Radiation and environmental protection

Radiation and environmental protection - 609

Optimization of 4H-SiC Neutron Detector Efficiency for Enhanced Border and Port Security

Bernat Robert1, Ivana Capan2, Vladimir Radulović3, Klemen Ambrožič3, Luka Snoj3, Zoran Ereš2, Takahiro Makino4, Takeshi Ohshima4, Adam Sarbutt5, Željko Pastuović5

1Ruđer Bošković Institute Department of Occupational Safety and Health, Fire and Radiation Protection, Bijenička cesta 54, 10000 Zagreb, Croatia

2Rudjer Boskovic Institute, Bijenicka cesta 55, 10002 ZAGREB, Croatia

3Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

4National Center for Scientific Research “DEMOKRITOS” Institute of Nuclear and Radiological Sciences and Technology, Energy and Safety Research Reactor Laboratory, PO Box 60228, 15310 Agia Paraskevi, Attiki, Greece

5Australian Nuclear Science and Technology Organisation - ANSTO Communications Government & Public Aff., New Illawarra Road, Lucas Heights, 2234 Sydney NSW, Australia


Research of wide bandgap semiconductor materials is increasing as the world is experiencing a shortage of helium-3. This paper continues the work previously reported within the E-SiCure project (Engineering Silicon Carbide for Border and Port Security), co-funded by the NATO Science for Peace and Security Programme, and focuses on the development of a converter for efficient detection of thermal neutrons. Simulation of the optimal thickness for thermal neutron converter have been performed using two following tools: Monte Carlo N–Particle Transport Code (MCNP) and Stopping and Range of Ions in Matter (SRIM). For the neutron converter material, we have used 6LiF and 10B4C of few different thicknesses on glass and aluminium substrate. Several different active area sizes of the detector based on 4H-SiC have been tested. Neutron irradiations were carried out at the Jožef Stefan Institute’s TRIGA Mark II reactor. For the calibration of the detectors we have used several different alpha sources. Data collected by simulations have been compared with those obtained from the irradiations in the TRIGA Mark II reactor.

09.09.2020 11:50 Radiation and environmental protection

Radiation and environmental protection - 610

Natural Radionuclides in Modern World and Basic Safety Standards

Helena Janžekovič

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


One of the main topics in nuclear industry is understanding risks associated with nuclear fuel cycle, i.e. mining and milling of uranium or thorium ore, production of fuel, operation of nuclear power plants, reprocessing, decommissioning and managing radioactive waste and spent nuclear fuel. In particular risks associated with radioactivity are in the focus of nuclear safety. These risks to human beings and the environment are controlled basically on a case-by-case basis, i.e. while some countries decided to have some parts of the nuclear fuel cycle, i.e. Australia, others decided not to have any element of the cycle on their territory. International safety standards, e.g. IAEA standards, exist in order to achieve common understanding of the risks.
The economy of nuclear power plants is driven among others by comprehensive safety rules in order to assure nuclear safety. In this respect, it is very important that designers, operators, regulators, radioactive waste agencies and other stakeholders understand safety standards related to ionizing radiation. These safety rules are widely based on recommendations given by International Commission on Radiological Protection. In the European Union (EU) three basic directives stream nuclear safety and radiation protection, i.e.:
1. Council Directive 2011/70/EURATOM on safe management of spent fuel and radioactive waste,
2. Council Directive 2009/71/EURATOM on nuclear safety and its amendment from 2014,
3. Council Directive 2013/59/EURATOM on basic safety standards for protection against the dangers arising from exposure to ionising radiation.

They form harmonized set of requirements, some of them are technical. In particular, the Council Directive 2013/59/EURATOM is very technical, e.g. with more than 1100 technical parameters. It is based on the ICRP 103. Among many novelties it opens a door for controlling risks associated with natural radiation sources.

As radiation protection community is very often confronted with an assumption stating that natural radiation sources do not pose any risks to the human beings it might be appropriate to analyse where natural radiation sources enter in the everyday life of a typical Central European State. The article gives the systematic overview on a required control over natural sources associated with building materials, radon and thoron, industries processing naturally-occurring radioactive materials and contaminated areas due to past activities among others. The international experiences based on open literature is given including experiences with consumer products such as radioactive pendants. Challenges associated with these issues regarding free movement of goods in the EU are also discussed.

09.09.2020 12:10 Radiation and environmental protection

Radiation and environmental protection - 611

Analysis of EURDEP radiation levels causes by Chernobyl forest fire

Marco Sangiorgi1, Miguel Angel Hernandez Ceballos2, Franck Wastin3, Marc De Cort1

2EC, Directorate-General Joint Research Centre Institute for Energy Safety of Present Nuclear Reactors Unit (SPNR) Plant Operation Safety, PO Box 2, 1755 ZG Petten, Netherlands

3European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands


Due to abnormally hot, dry and windy weather, forest wildfires broke out in Ukraine about 3/5 April 2020 in a territory still heavily contaminated by the 1986 Chernobyl nuclear accident. These wildfires reached the exclusion zone and the surrounding environment (about 1 km) of the nuclear power plant from April 8, 2020 and they are up to now the biggest fires ever recorded in the Chernobyl exclusion zone.

This event was of major interest due to the fact that since 1986, the forests have accumulated radioactivity, mostly concentrated in the wood and upper soil layers and, because of the fire haze these particles could be resuspended into the air and transported over long distances. Indeed the press reported detections of a small increase of caesium-137 concentrations in the air in Kiev, Germany, Norway, Greece and other countries. The EURDEP network also detected a slight increase of gamma dose rate with respect to the background in the proximity of the fire.

This paper analyses the ambient gamma dose rate measurements collected by the EURDEP network during those days and, matches them with the results of a calculation using the JRODOS software, which predicts aerosol resuspension caused by wildfires, transport through the atmosphere, further deposition to the ground and potential dose rates.

10.09.2020 11:10 Education and training

Education and training - 1604

How to train newcomers about nuclear safety

Andrej Stritar Stritar1, Igor Jenčič2, Tomaž Skobe2

1Retired, nn, nn, Slovenia

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


One of the most used terms in nuclear industry is nuclear safety. There are several simple definitions of it published by different authors and organisations. Each nuclear professional believes that he or she very well understands its meaning. However, when it comes to training of the newcomers about it, it becomes a challenge! Where to start, what to include, how much in detail should we go?
The Nuclear Training Centre at Josef Stefan Institute in Ljubljana is performing training of future Krško nuclear power plant operators. The initial, theoretical part lasts about six months and includes all needed areas like atomic and reactor physics, thermal-hydraulics, electrical engineering and radiation protection. There is also a module on nuclear safety included towards the end of the training course. This year it was decided to renew that module completely, as previous training materials were becoming outdated, divided into several sub-modules and were never reasonably consolidated into one comprehensive training module.
The new training module on nuclear safety starts by explaining the basic concepts: what is hazard, what is risk, what are the main hazards nuclear power plant can cause. This is followed by explanation of main principles used in nuclear industry, which are mainly based on IAEA Safety Fundamentals. In the next step, the trainees should learn how the nuclear power plant could come to the situation, where the core starts to melt and radioactivity starts to spread into the environment. They have to learn about all potential scenarios leading into the accident in the pressurized nuclear power plant. They have to understand what would happen if there is too much or too little heat removal from the core, if the primary coolant is lost or if external events like earthquakes or floods happen on the site. Here belongs also the explanation of physical phenomena during the severe accidents. The understanding of dangers and potential bad scenarios in the nuclear power plant is the basis for learning about design bases used during the construction. Special emphasis is given to the explanation of different NPP states, including design extension conditions introduced after Fukushima accident. At that point, it is the time to explain deterministic and especially probabilistic safety analyses. That leads the training also towards the explanation of Safety Analysis Report and Technical Specifications.
All the above was related to the period before the operation of the NPP. Before starting with the operation, the concept and importance of the safety culture is emphasised, supported by numerous examples. In the next chapter, it is explained about reactor protection system and safety systems, those parts of the NPP, which are not necessary for production of electricity, but are very important for safety. The importance of plant procedures, safety indicators and corrective actions follows. The national and international regulatory system is explained in the continuation. Finally, the on-site and off-site emergency preparedness and response arrangements are summarized. It is finished with the description how the response to the severe accident with radioactivity releases would look like on site and in the whole country. It is emphasised that the whole training module is oriented towards prevention that anything like that would ever happen.
In the last chapter three main nuclear accidents, Three Mile Island, Chernobyl and Fukushima are described.

10.09.2020 11:30 Education and training

Education and training - 1605

20 Years Of The NEK Full Scope Simulator

Matjaž Žvar

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia


In March 28th 1979 the mechanical failures at the Three Mile Island were compounded by the initial failure of plant operators to recognize the situation as a loss-of-coolant accident due to inadequate training. In the following years, the outcome, among other things, was a demand for every utility to have a plant specific simulator for operator training. The Slovenian Nuclear Safety Administration issued their simulator demand to NEK in April 1995. The first training session on the simulator was performed on April 17th in 2000 and since then the simulator has been used on daily bases to improve operator knowledges, skills and performances.

This was also the first full scope simulator with the capability, what was unique at that time and still is, to simulate Beyond design basis accidents. That makes it very suitable for emergency preparedness drills, because core meltdown can be simulated, even with a containment breach. After the 2017 simulator upgrade, fuel in the spent fuel pit can be melted as well. The tool used to simulate beyond design bases accidents is the Modular Accident Analysis Program – MAAP 5 which interfaces with the simulator model. Nowadays, even severe accident mobile equipment is implemented in the simulator.

Besides operations training, there is a variety of ways how to use the simulator. It can be used for implementation of modifications before plant implementation for their testing or for just-in-time training for infrequent performed evolutions or for procedure development. The Pressurized Water Reactor Owners Group (PWROG) used the NEK simulator in 2019 to develop the new set of the Severe Accident Management Guidelines, incorporated with a completely new approach of the guideline’s usage. It’s even used for popularization of the nuclear energy for our visitors.

In all of these years the simulator has been actively participating in the increased reliability and stability of the electricity production and in achieving NEK’s vision to be a worldwide leader in nuclear safety and excellence.

10.09.2020 08:30 Nuclear fusion and plasma technology

Invited lecture - 101

The international platform WEST to prepare ITER operation and beyond

Jérôme Bucalossi

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


Power exhaust is a challenge for ITER operation and a potential showstopper on the roadmap towards fusion energy. Reliable power exhaust requires a thorough integration of physics and technology. For this purpose, CEA built the international platform WEST tokamak, which is the transformation of the French Tore Supra facility into a diverted tokamak fully tungsten environment with a set of novel actively cooled plasma facing components (PFC). One of WEST's key missions is to minimize the risks for the manufacture and operation of the ITER divertor. WEST also fills the gap on long pulse tokamak operation within the European fusion program. It offers an integrated tokamak environment for ITER, also for testing and qualification of components in future fusion devices. The WEST research program, imbedded within the EUROfusion program, is carried out in close collaboration with ITER Organization and ITER partners, in particular, China, India, Korea and US.
The assessment of plasma facing, component (PFC) performance and lifetime under relevant power fluxes and particle fluence are the central thrust of the program. Other issues including operation at high radiated fraction in compact divertor geometry, demonstration of detachment control over long pulse, exhaust physics at large aspect ratio and operation in double null are key topics which are also tackled in the perspective of the fusion reactor.
In WEST, 16 MW of RF power (9 MW of ICRH and 7 MW of LHCD) provide relevant heat and particle load conditions on the divertor over duration up to 1000 s. The lower divertor is equipped with ITERgrade components in order to test them and gain operational experience in advance of ITER operation. A new ultra high-resolution infrared thermography system is under deployment, fully covering the lower divertor with millimeter spatial resolution. In addition, sets of Langmuir probes (foreseen in ITER) together with thermocouples and Fiber Bragg Grating are embedded in the divertor target. Finally, calorimetric sensors offers real time monitoring of PFC cooling circuits allowing for accurate power balance. Such a powerful set of diagnostics allows developing sophisticated protection schemes in view of ITER divertor operation.
In this paper, we report substantial progresses recently achieved, particularly the achievement of 1 minutes pulses, steady-state divertor heat fluxes in the range of 5 MW/m2 and transitions towards high confinement regimes (H-mode). Experiments with the full ITER-grade divertor will start in autumn 2021, with the ultimate target of running 1000 s plasmas with the heat fluxes expected on ITER PFC (i.e. 10 MW/m2).

7.09.2020 17:20 New reactor design and small modular reactors

Invited lecture - 104

NuScale SMR Technology – Smarter, Safer, Cleaner, Cost Competitive and US NRC-approved

Scott Rasmussen


NuScale’s SMR design is the first ever to receive U.S. NRC approval, a significant milestone not only for NuScale, but also for the entire U.S. nuclear sector and the other advanced nuclear technologies that will follow. NuScale’s innovative and unique SMR design has unparalleled safety and reliability features, can be fully factory-made, and offers scalable power based on need – an unprecedented capability in the nuclear energy industry. A 12-module NuScale SMR Plant generates up to 720 MWe – a substantial amount of power at a fraction of the cost and size of traditional gigawatt plants.

7.09.2020 17:00 New reactor design and small modular reactors

Invited lecture - 103

Holtec’s SMR-160: Safe, Secure, Reliable, Flexible, and Economical Clean Energy to Support the World’s Energy Needs

Scott Rasmussen


SMR-160, developed by Holtec International (USA), is a small modular reactor designed to produce 160 megawatts of electricity using low enriched uranium fuel. SMR-160 is intended to serve as a distributed energy source that dispenses with the need for expensive high capacity transmission lines over long distances, making the electric grid more resistant to natural disasters or acts of sabotage.

Informed by over six decades of lessons learned from reactor operations, SMR-160 is designed to be an unconditionally safe reactor, which means it will not release radioactivity regardless of the severity of the natural or manmade disaster. Every conceivable catastrophic event – severe cyclones (hurricanes or typhoons), tsunamis, flood, fire and crashing aircraft – has been considered, with appropriate features incorporated in the design of SMR-160 to ensure that it will withstand these events without releasing radioactivity or pose any risk to public health and safety. In other words, SMR-160 is an industrial installation from which one will safely walk away in the wake of an unexpectedly severe natural disaster or act of sabotage, letting the plant’s innate defenses look after the reactor’s wellbeing.

Because SMR-160 is walk away safe, it can be sited next to population centers without any threat to the local environment or populace. It is as benign to its host locale as a cotton mill or a chocolate factory. Placing SMR-160 close to cities and towns will reduce transmission losses and enable the plant’s workers to live in the local community. A SMR-160 installation takes up less than 4.5 acres of land; this is a fraction of the land area required by other types of power plants.

7.09.2020 16:40 New reactor design and small modular reactors

Invited lecture - 102

75 years of innovation. Energy and non-energy solutions for better life

Anton Moskvin


This year Russia is celebrating 75-year anniversary of the nuclear industry. Nuclear technology in Russia had passed a long way from exclusively science field to multifunctional solution that can be applied in various fields. Today Rosatom is one of the nuclear industry leaders and ready to provide required energy solutions.

Rosatom is a unique vertically integrated enterprise that covers the complete nuclear fuel cycle from uranium mining throughout NPP construction and operation to back-end and decommissioning. Rosatom also develops its business and research in non-energy sectors such as research reactors, radiation technologies and nuclear medicine.

Being a global player with many projects in Russia and overseas, Rosatom keeps leading role in NPP project development and providing Russia and overseas partners with stable and clean electricity supply. Currently Rosatom holds approximately 70% of the global NPP construction market of overseas projects. Now Rosatom has 36 units in 12 countries in its overseas projects implementation portfolio. For the past 14 years, Rosatom has commissioned 15 NPP units both in Russia and abroad. NPP construction projects are being implemented in Finland (VVER-1200), Hungary (VVER-1200), Turkey (VVER-1200), Belarus (VVER-1200), Egypt (VVER-1200) and Bangladesh (VVER-1200).

VVER-1200 reactor is Rosatom’s safe and economically efficient flagship reactor technology. It is a time-tested and highly referential energy generating solution of generation III+, designed in strict compliance to post-Fukushima safety requirements. Generation III+ nuclear power plant with VVER-1200 technology combines successful experience in NPP operation with cutting-edge safety standards, while meeting the most stringent requirements. Russian-designed VVER reactors have successfully undergone international stress testing. VVER reactors are operated in a number of the EU countries, including Finland, Czech Republic, Hungary, Slovakia and Bulgaria.

Rosatom is ready to provide countries an access to the whole range of non-energy nuclear applications. Research reactor-based Centers for Nuclear Science and technology (CNST) projects are one of Rosatom’s key products to the global market. These centers are an innovative solution that formed from a research reactor and a set of nuclear research laboratories as well as nuclear medicine center and multipurpose irradiation center. CNST can be applied for a wide variety of scientific and industrial purposes, starting from educating students and training HR to radiopharmaceuticals production and other high-end activities.

Rosatom has built 122 research reactors in the past 70 years and is currently operating 20% of all research reactors worldwide. Rosatom designed and built research reactors are operated in Czech Republic, Hungary and Poland.

Developing a Center for Nuclear Science and Technology project helps countries with advanced nuclear energy program to maintain a high level of competence of personnel working in the field of peaceful atom. It also allows providing scientific and technical support of NPP operation, experimental works in the field of radiation materials science, reactor physics and metrology, reactor tests and irradiation of samples, products and equipment for the needs of nuclear power. To ensure that the country will fully develop competencies in the field of nuclear technologies Rosatom provides support in the systematic development of the national nuclear program: from stakeholders engagement in strategic planning for nuclear program development and optimizing of the CNST configuration to the assistance in integration into international teams of professionals.

Today we see demand for such comprehensive solution from the global market; many countries from various regions are interested in implementing CNST projects. In 2018, such project was launched in Bolivia. Interest in the Centers development is growing among countries from different regions including Africa, Asia and Middle East.

Rosatom boasts solid experience not only in research reactor design, construction or operation but also in modernization, upgrade and lifetime support. One of the options for partner countries is integrated upgrade of the nuclear island equipment, such as replacement of circuit equipment, reactor vessel, core/reflector etc. For example, Dalat research reactor in Vietnam was reconstructed and upgraded from TRIGA Mark II reactor. Rosatom may also assist in conversion to low-enriched or alternative fuel and supply new experimental devices and components. Rosatom provides solutions for upgrade of the I&C systems of Russian and foreign design: design and upgrade of control and protection systems, devices, software and hardware for control systems safety and normal operation, development of computer-driven control or operator support systems, emergency control console supply as well as design supervision of systems and equipment during fabrication, research reactor installation and further operation.

10.09.2020 09:10 Nuclear fusion and plasma technology

Nuclear fusion and plasma technology - 517

Deposition of the ITER Vacuum Vessel dust inside the pressure suppression system during a Loss of Coolant accident: experimental and numerical analyses

Miriam Ibba1, Roberta Lazzeri1, Alessio Pesetti2, Andrea Marini1, Flavio Parozzi3, Biswanath Sarkar4, Marco Olcese4, Donato Aquaro1

1University of Pisa, Via Montebello di Mezzo 17, 19020 Bolano, Italy

3Research on energy systems - RSE S.p.A, via r. Rubattino, 54, 20134 milano, Italy

4ITER Organization, Cadarache Centre, Building 519, 13108 St. Paul lez Durance, France


During the operation of a nuclear fusion reactor, an amount of dust is accumulated in the vacuum vessel due the erosion phenomena produced by the plasma. This dust is mainly compound of tungsten and beryllium produced by the erosion of the divertor and of the first wall, respectively.
In the case of accidental events in the vacuum vessel, like Loss of Coolant Accident, above all the beryllium dust (greatly as beryllium oxide) is transported to the pressure suppression system and to draining system. The dust deposition inside of these components and the necessary activities of decontamination are considered important issues from the safety point of view.
At the Department of Civil and Industrial engineering of the University of Pisa (Italy), an experimental and numerical research program concerning the deposition of the dust in a reduced scale pressure suppression system is carried out, funded by ITER Organization.
Preliminary laboratory scale experiments have been carried out in order to determine a best simulant of beryllium dust to use in the deposition/decontamination tests performed in the experimental rig. In fact, the beryllium dust is a dangerous element from the health point of view.
Al2O3 dust with suitable granulometry has been demonstrated to be a good simulant of beryllium oxide.
Experimental campaign is being carried out to analyse the deposition behaviour of Al2O3 dust injected in the water with a flow of air and steam.
The research program has several objectives. To determine :
- the dust which is deposited on the horizontal wall and on the lateral wet and dry walls
- the dust which is entrained by the water during the discharge
- the dust which is entrained by the air flow . This quantity permits to calculate the scrubbing efficiency of the apparatus.
- the efficiency of the cleaning system which operates after the water discharge for the decontamination of the component
The ranges of the test parameters are the following:
- Water mass : 2-3 m3; water head: 0.5-0.85 m; dust mass flow rate: 0.6-1.6 g/s; steam mass flow rate: 10-50 g/s (T=150°C, 1.5 bar); air mass flow rate: 100-500 m3/h.
The previous data are consistent with the use of a reduced scale experimental rig (1/22 scale factor).
Numerical simulations have been performed in order the estimate the sensitivity of the different parameters (water head, air, steam and dust flow rates, dust granulometry) on the scrubbing efficiency.
The simulations have been performed by means the ECART code (ECART User's Manual – F. Parozzi – RSE- spa) which calculates the transport of aerosol substances throughout pipes or plant components.
A simplified model of the experimental set up, shown in Figure 3, has been implemented. Aerosol transport mechanisms (growth, agglomeration, deposition, scrubbing and resuspension) are accounted in detail by models. The dust granulometry has been considered by means of a discretized particle size distribution. The interaction between the fluid and the walls, as well as the thermal conduction within the wall materials is also calculated. It is possible to evaluate pool scrubbing effects in the simulations in which is injected also an additional air mass flow rate (through the junctions J8 in Figure 3).
The numerical results been compared with the experimental ones. A good agreement has been obtained in terms of scrubbing efficiencies.

10.09.2020 09:30 Nuclear fusion and plasma technology

Nuclear fusion and plasma technology - 516

The role of the neutrals in the parallel filamentary transport in the tokamak SOL

Jernej Kovačič1, Stefan Costea2, Tomaz Gyergyek1, Tsviatko Popov3, Roman Schrittwieser4

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

3Scientific Research Department (NIS), 5 James Boulcher Blvd., 1164 Sofia, Bulgaria

4University of Innsbruck Institut of Ion Physics Plasma Department, Techniker str. 25, A-6020 Innsbruck, Austria


The heat and particle loads on the plasma-facing-components (PFCs) in the tokamaks are one of the main limiting factors in the design of the future fusion power plants [1]. Unfortunately, the modelling of the plasma scrape-off-layer (SOL) in a tokamak is very difficult due to a variety of processes and mechanisms affecting the behaviour of the particle, momentum and energy transport, such as elastic and non-elastic collisions, finite cyclotron radius, magnetic field curvature, strong local electric fields, etc. Consequently, fluid modelling is often employed, where these processes are taken into account through averaged transport coefficient. However, the SOL plasma does not fulfil the conditions needed for fluid treatment due to long mean free paths of charged particles and as such needs an approach that can take into account divergence of plasma from the thermodynamic equilibrium.
In recent years we have developed a fully kinetic model of an MST tokamak using the massively parallel BIT1 code [2], which we have modified in order to be able to simulate a filamentary source of plasma. Furthermore, a new sink model has been developed for the code to mimic the radial loses. The code’s main feature is its’ rich assembly of collisions implemented into the Monte-Carlo collisional module, ranging from Coulomb collisions to charge-exchange, ionization, excitation, etc. This allows us to study the effects of the neutrals on the transport properties in the SOL.
In our study we have focused on the role of neutrals that migrate into the SOL due to recycling on the divertor plates and on the chamber walls and due to fuelling/puffing at the divertor and at the outer midplane. We have obtained time dependant profiles of macroscopic quantities, detailed energy distribution functions, spectroscopic data and energy fluxes. The results show how the increased collisionality due to interaction with neutrals inhibits the parallel transport and leads to increased transport in the radial direction. This could be a partial explanation to the hot question of density shoulder formation in the SOL, e.g. [3]. We found out, that these effects are dependent on the charged particle energy distributions, which define the position of the ionization fronts, the types of collisions, etc., proving that kinetic approach is indeed needed.

10.09.2020 09:50 Nuclear fusion and plasma technology

Nuclear fusion and plasma technology - 518

Thermal loading of DEMO Breeding Blanket during maintenance conditions

Martin Draksler1, Jakob Justin2, Christian Bachmann3, Boštjan Končar1

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

3PPPT, PMU, EUROfusion Consortium, Boltzmannstrasse 2, 85748 Garching, Germany


Prior to removal of In-Vessel Components (IVC), the vacuum vessel is filled either with dry nitrogen or with humidity-controlled air at absolute pressure of 1 bar. Then the active cooling loop of the IVC is deactivated and drained. Due to the decay heat generated by the generated radioactive isotopes inside the IVC material, the IVC is heated internally. The established natural convection inside the Vacuum Vessel (VV) transfers the generated decay heat to the cold IVCs and VV wall, which remain actively cooled to room temperature. As the removal of the IVC by remote handling (RH) is foreseen one month after the DEMO reactor shutdown, the conservative analysis assumes the calculation of the steady-state condition using the constant decay heat value at one month after the shutdown. The maximum temperatures in the removed IVC component must not in any way exceed the limiting temperature of 150°C, thus it is essential to predict its thermal loading already in the design phase.

This study evaluates the thermal loading of DEMO breeding blanket segment without active cooling during the remote maintenance scenario. The aim of the study is to assess the importance of the radiation heat transfer comparing to the heat transfer by natural convection. The obtained results are compared with the results of our previous study where the radiation heat transfer was neglected. Steady-state CFD analysis with ANSYS CFX code 19.3 [3] has been carried out to assess the temperature distribution of the in-vessel components and redistribution of thermal loads during the maintenance scenario in DEMO tokamak.

08.09.2020 17:00 Severe accidents

Severe accidents - 1809

The new MERELAVA facility for ex-vessel corium flooding

Christophe Journeau, Viviane Bouyer, Arthur Denoix, Pascal Sauvecane

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


CEA has built MERELAVA, a new facility in its PLINIUS platform to study the water flooding of prototypic core-concrete interaction melts containing significant fractions of steel. 60 to 80 kg of corium-concrete mixture can be molten by thermitic reaction before water flooding. The heat flux extracted by boiling water is assessed through 3 complementary measurements (composition, mass flow and temperature of exhaust gases; heat extracted at the condenser; volume of condensates).
This new facility will be described with its capabilities. A first experiment is devoted to the flooding an oxidic melt. Further experiments will deal with increased steel fraction in the melt in order to assess its effect on water ingression. Then, the facility will be upgraded to carry out experiments with sustained induction heating.
Acknowledgement: This work is part of the MIT3BAR project (MITigation of 3rd BARrier failure) carried out within the framework of the RSNR initiative funded by the French Government ‘Investments for the Future’ program and managed by the French National Research Agency (ANR-10-RSNR-01). The support of EDF and Framatome is gratefully acknowledged as well as the work and efforts of the PLINIUS experimental team.

08.09.2020 17:20 Severe accidents

Severe accidents - 1810

Premixed layer formation modelling in stratified configuration

Janez Kokalj, Mitja Uršič, Matjaž Leskovar

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


A hypothetical severe accident in a nuclear power plant has the potential for causing severe core damage, including core meltdown. If the hot melt comes in contact with the coolant water, the internal energy is rapidly transferred, which can result in a steam explosion. Considering the amount of thermal energy, initially stored in the liquid corium melt at about 3000 K, this phenomenon can jeopardize the integrity of the containment and can cause damage to the systems inside. Consequently, the possibility of a radioactive leakage presents danger for the environment and general public safety. Similar explosion phenomena can be a concern in some industrial processes, such as foundries and liquefied natural gas operations or in certain volcanic activity where water is present.
In nuclear safety, the steam explosions are mostly analysed in the melt jet-coolant pool configuration. Stratified configuration was believed to be incapable of producing energetic fuel-coolant interaction. However, the results from recent experiments performed at the PULiMS and SES facilities (KTH, Sweden) 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 steam explosions.
The purpose of our research was to improve the knowledge, understanding and modelling of the fuel-coolant interaction phenomena in the stratified configuration. In the paper, a model for the premixed layer formation, based on the visual observations and some mechanisms from the literature, will be presented. The developed model was implemented in the MC3D code (IRSN, France) and validated against the experimental results. With a short analysis, we will show that the model is capable of describing the premixed layer formation.

08.09.2020 17:40 Severe accidents

Severe accidents - 1811

Experimental investigations on the retention of soluble particles by pool scrubbing

René Vennemann1, M Klauck2, Hans-Josef Allelein3

RWTH Aachen University Institute for Reaktor Safety and Reactor Technology, , 52062 Aachen, Germany

3Institute for Reactor Safety and Reactor Technology, Eilfschornsteinstraße 18, 52062 Aachen, Germany


In the late stage of a severe loss-of-coolant accident, the pressure in the containment building of a nuclear power plant could rise beyond the design limits and thus endanger its structural integrity. Therefore, to avoid pressure failure, it may be necessary to per-form controlled venting of the containment. During the event of the accident, a large amount of fission and activation products is released into the containment as airborne particles (aerosols). These particles are filtered during the venting process, usually with the help of wet filters, in order to keep risks to the surrounding environment to a mini-mum. Consequently, the knowledge of the retention processes in a water reservoir (pool scrubbing) is of central importance for such filtered containment venting systems (FCVS) and for reactor concepts in which water reservoirs are used for pressure reduc-tion (e.g. condensation chamber of a BWR).
Investigations on pool scrubbing are carried out in the SAAB test facility at the Jülich Research Centre. SAAB is a unique large scale facility with the ability to perform a great variation of experiments using various measurement tools. The influence of nu-merous parameters, such as the height of the water pool, solubility and concentration on the retention capacity, is investigated by means of single effect studies on both insoluble and soluble particles. This paper gives a detailed overview over the facility and includes the results of the first test series with soluble particles (incl. CsI).

08.09.2020 18:00 Severe accidents

Severe accidents - 1812

Capture of gaseous radionuclides in porous Metal-Organic Framework

Maeva Leloire1, Philippe Nerisson2, Olivia Leroy3, Laurent Cantrel4, Christophe Volkringer5, Thierry Loiseau2

1IRSN - Institut de radioprotection et de sureté nucléaire, Nuclear Safety Division , BP17, 92262 Fontenay-aux-Roses Cedex?, FRANCE, France

3Institut de Radioprotection et de Sureté Nucléaire (IRSN) Centre d’ Etrudes de Cadarache, Cadarache B.P 3, Batiment 702, F-13115 Saint Paul-les-Durance CEDEX, France

4Institut de Radioprotection et de Sureté Nucléaire, 31, avenue de la Divison Leclerc, 92260 Fontenay Aux Roses, France


Isotopes of fission products (FP) such as iodine and ruthenium, usually 131I, 103Ru and 106Ru, are usually produced in large amount by nuclear fissions. After a nuclear accident, these elements can be quickly disseminated since they may generate very volatile species such as molecular iodine (I2) or ruthenium tetroxide (RuO4). In order to limit fission products dispersion, filters made of porous materials are used or could be used in nuclear plants for FP mitigation. However, such pure inorganic porous solids exhibit many limitations in the case of a nuclear accident, especially in the presence of poisoning species (e.g. NOx, H2O, COx) or for the capture of large molecules such as RuO4.
Based on these limits, a relative new class of porous materials called Metal-Organic Framework (MOFs) could be an effective substitute. Indeed, MOFs are hybrid materials, composed of inorganic clusters linked to each other by organic ligands. This low-dense organization allows important porosity and records in specific surface (up to 7000 m2.g-1). Another advantage of MOFs over zeolites and activated carbons, is the possibility to easily functionalize their frameworks through the organic sub-network. This strategy allows to adjust the size of the pores, as well as to attach specific chemical groups improving the capture of the desired molecules.
So far, the efficiency of MOFs for the capture of radioactive gaseous species is not well documented and many questions remain. For example, this family of materials was never tested for the capture of volatile RuO4. Furthermore, the shaping and the contribution of functional groups was not yet examined for the capture of iodine.
Therefore, this work deals with the capture of volatile I2 and RuO4 in MOFs. In this communication, we will highlight the importance of MOF functionalization and shaping for the capture of gaseous radionuclides; especially under accidental conditions.
The capture of the different species was realized in homemade installations, allowing constant flow of I2 and RuO4, and the quantification of species trapped within the porous structure.
The capture of gaseous molecular iodine I2 is studied in the isoreticular series of stable Zr based MOFs called UiO-66, UiO-67, and UiO-68. This particular family of compounds gives an access to a wide range of functional groups (-H, -Cl, CH3, -NH2, etc.) and pore diameters (up to 17 A). Especially, we showed the formation of charge transfer complexes between amino group attached to the MOFs and iodine, leading to very high uptake (more than 1 gram of I2 per gram of MOFs). This particular interaction was characterized by a set of techniques (XRD, Raman spectroscopy, etc.) and confirmed by modelling calculation.
For the capture of RuO4, we focused our work on MOF-808-(Zr), which exhibits a very good stability under drastic conditions, as well as very large cages (size 18 A) able to trap big molecules such as RuO4. Indeed, this MOF reaches a decontamination factor up to 200, and contains 33 %wt Ru after trapping. When RuO4 is trapped in MOF, it decomposes into its most stable and nonvolatile oxide RuO2. This last species remains permanently trapped into the MOFs cage as nanoparticles. These results were, inter alia, confirmed by transmission electronic microscopy.

08.09.2020 18:20 Severe accidents

Severe accidents - 1813

Severe Accident Simulations Dedicated to the SAMG Decision-Making Tool Demonstration

Piotr Darnowski1, Ivica Basic2, Ivan Vrbanic2, Maciej Skrzypek3, Janusz Malesa3, Ari Silde4, Jarno Hiittenkivi4, Piotr Mazgaj1, Luka Štrubelj5

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

2APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia

3National Centre for Nuclear Research, ul. Andrzeja Sołtana 7, Otwock-Świerk, Poland

5GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia


The paper presents severe accident simulations performed to generate a database of plant states dedicated for use with Severe Accident Management Guidelines Decision Making Tool (SAMG DM). The software is being developed in the framework of the NARSIS Horizon-2020 Research Project. It is intended to be a supporting tool for the SAMGs implementation, Emergency Preparedness and selection of Severe Accident Management strategies. Simulations were performed with MELCOR 2.2 integral computer code for generic Nuclear Power Plant with Gen-II Pressurized Water Reactor, being representative for European nuclear fleet. The database covers results for the most important phenomena and parameters in the RCS and containment, for the in-vessel phase of different accidents. Two general types of scenarios were considered, low-pressure sequences (LB-LOCA) and high-pressure sequences (SBO). Several variants were included, for low-pressure CL and HL breaks with and without recovery, for high-pressures, it covers seal LOCA, SG tubes rupture, SL rupture, different SV positions and recovery setup. The database applicability was evaluated, its limitations and areas of application were estimated. The landscape of scenarios in the database creates phase-space of plant states being large enough for DM tool implementation and demonstration.

09.09.2020 14:00 Materials in nuclear technology

Materials in nuclear technology - 1108

In-Situ organic acid generation to catalyse Solketal Production from Glycerol using ionizing irradiation from a TRIGA Reactor

Arran Plant1, Malcolm Joyce1, Vesna Najdanovic-Visak1, Bor Kos2, Anže Jazbec3, Luka Snoj2

1Lancaster University, Department of Engineering, Bailrigg, Lancaster, LA1 4YW, United Kingdom

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

3Jožef Stefan Institute, Reactor Infrastructure Centre, Jamova cesta 39, 1000 LJUBLJANA, Slovenia


In this paper, the conversion of the renewable waste chemical, glycerol, will be described which utilizes ionizing energy from JSI’s TRIGA fission reactor. The end compound, solketal, is a desirable product due to its potential to be used as a renewable fuel additive or a green solvent. This research could help realise an integrated system for co-production of valuable chemicals feedstocks from organic waste using energy from fission reactor systems, alongside the production of heat and power.
Previous comparisons between mostly neutron-irradiated (n+??) and gamma (?) irradiated glycols showed no difference between the detected radiolysis components and little difference in their respective yields. The limiting factors for the synthesis of solketal were theorised to be the availability of the radiolysis-generated reagent, acetone and the availability of acidic compounds to act as a catalyst. New samples have been irradiated with which to explore factors associated with vial types, dose rates and ternary mixtures. The radiochemical products have been identified, confirmed and quantified using Gas Chromatography-Mass Spectrometry (GC-MS) techniques together with calibrated analytical standards.
Glycerol-acetone-water mixtures gamma-irradiated to doses of 50 kGy showed significant increases in solketal (2,2-Dimethyl-1,3-dioxolane-4-methanol) conversion compared to the neat glycerol, showing a 1.8% mass conversion from glycerol. It is thought that with the accompanying concentrations of acetic acid, derived from acetone radiolysis and lower dose rates allows a higher yield of solketal to be formed during irradiations. With many publications focussed on using traditional catalytic methods, this is currently considered to be the only recorded work which has utilized irradiation or ionizing radiation to promote the conversion of glycerol to solketal. Collaborative research with MCNP simulations will also be described which explores the feasibility of the incorporation of a radiolysis loop. This has been achieved by adapting existing MCNP models for both the JSI’s TRIGA reactor and the commercial NPP at Krsko, Slovenia.

09.09.2020 14:20 Materials in nuclear technology

Materials in nuclear technology - 1109

Preliminary analysis of the creep and ageing influence during SBO accident

Salvatore Salvo Cancemi1, Rosa Lo Frano1, Piotr Darnowski2, Piotr Mazgaj2, Riccardo Ciolini1

1University of Pisa, Department of Civil and Industrial Engineering, Largo Lucio Lazzarino 2, 56124 Pisa, Italy

2Warsaw University of Technology, Pl. Politechniki 1, 00-661 Warsaw, Poland


Severe accidents involve a set of difficult phenomena to understand and consequences to predict because of complex interactions occurring at high temperatures between materials. Complexity (of the heat transfer behavior) increases further during the molten relocation of corium to the lower head of the reactor pressure vessel, such that the resulting thermal loads may threaten the strength of the vessel wall.
In this framework and with reference to the SBO (station black out) accident, the role played by the creep, and also the influence of aging on it (in terms of reduction of bearing capacity of the vessel lower head) are studied. In doing that an external coupling between MELCOR and MARC codes is performed: the output temperatures and pressures obtained from the first code represent the input for the thermo-mechanical simulations carried out with the latter one. Moreover, ageing phenomena are implemented through degraded mechanical material properties.
A suitable three-dimensional model allows to analyze the nonlinear behavior of the reactor vessel undergoing thermo-mechanical creep in SBO conditions. The deformation in the structure is calculated using nonlinear material propriety, such as the reduction of Young’ modulus and yield strength of carbon steel alloy.
The obtained results show that the thermo-mechanical loads are responsible of deformation of the vessel which develops and increases as the transient progresses. They also highlight the creep deformation process appears where the maximum temperature is located.

Keywords: Safety, SBO, creep, ageing, simulation, failure

09.09.2020 14:40 Materials in nuclear technology

Materials in nuclear technology - 1110

Microstructural evaluation of AISI 321H after supercritical water exposure

Daniela Marušáková1, Claudia Aparicio2

1Research centre Rez, Hlavni 130, 250 68 Husinec-Řež, Czech Republic

2Instituto Peruano de Energia Nuclear, Av. Canada 1470, San Borja, Lima 41, Peru


Heat-resistant stainless steel AISI 321H is main material used for internal components of nuclear reactors VVER 440 and 1000. New types of Generation IV reactors, specifically Supercritical water reactor (SCWR) can use some of materials which are already used in Generation II reactors, but corresponding experimental data are still needed. SCWR will operate at a pressure 25 MPa and core temperature from 350 to 625 °C. In that purpose, several cycles to supercritical water at 495 °C and 25 MPa took place in supercritical water loop (SCWL). Duration of each exposure was 500 h, 150 h and 1000 h. Scanning Electron Microscopy (SEM) with Electron Backscatter Diffraction (EBSD) in combination with X-Ray Diffraction (XRD) and Raman Spectroscopy (RS) were used to evaluate the microstructure of AISI 321H after the exposure to supercritical water. All these methods confirmed a thick non-uniform layer, which consists of magnetite crystals with thickness less than 1µm. Density of crystals increased after the second and third exposure. Crystallography of matrix have not changed after all exposures.

10.09.2020 11:50 Regulatory issues and legislation

Regulatory issues and legislation - 1404

EU environmental law as relevant to nuclear facilities

Ana Stanič

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


Since environmental impact assessments are mandatory for nuclear facilities an understanding of EU environmental law is key to ensuring the timely and on budget delivery of nuclear projects. This paper discusses the main provisions of EU environmental law concerning the construction, modernising and decommissioning of nuclear facilities as well as of storage and disposal of nuclear waste. Its aim is to provide legal, practical and strategic guidance to engineers, in-house counsel and the management of nuclear facilities on how to manage the EIA process including relations with the European Commission. In particular it will look at the recent decision of the Court of Justice of the EU regarding extending the life of nuclear facilities and the environmental considerations relevant thereto.

10.09.2020 12:10 Regulatory issues and legislation

Regulatory issues and legislation - 1405

Reviewing last Slovene Endeavours to Strengthen Safe Transport of Radioactive Material and a Quick Look over the Horizon

Janez Češarek

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


There haven’t been any traffic (road, rail, sea, air) accidents during transport of radioactive material in Slovenia in the past years. The total number of such shipments and the number of packages is relatively small, having in mind large European countries. Nevertheless, “a few thousands” packages with radioactive material with very different activities do require graded approach and precisely chiselling regulatory requirements, vigilance and human resources. It is wise and a matter of course to looking globally and having in mind others’ experiences and lessons-learnt from accidents with radioactive material which do occur from time to time. The best way is to look into such cases and prevent domestic re-occurrences. Both safety and security culture tend to act like a glue also for the issue of transport and its resilience when being challenged.
The backbone of the system (also) in Slovenia is to follow the European Agreement concerning the International Carriage of Dangerous Goods by Road (ADR) as well the requirements from the other modal regulation. This is enshrined through a direct link in the Act on Transport of Dangerous Goods. In addition, the “Nuclear Act, ZVISJV-1” gives a platform for requirements for transporting certain radioactive sources – to be a radiation practice – with all the consequential requirements and duties, imposed on the carrier. Transport of nuclear (fissile) material is another category which deserves particular attention. On the other hand, the International Atomic Energy Agency (IAEA) has produced a number of useful, transport-related documents and the current (2018 Edition) of the Regulations for the Safe Transport of Radioactive Material, known as “SSR-6”, has been clearly its flagship.
Avoiding to speak about a mature posture in this sphere, but a bunch of actions in the last couple of years has charted the path ahead. The Slovenian Nuclear Safety Administration (SNSA) established an informal group on safe transport of radioactive material back in 2017 – bringing together not only various public/governmental entities but also a handful of carriers and organisers of the carriages. One of such engagements has been the co-operation with the Slovene Police while elaborating (general) annual threat assessment for transports of potentially highly dangerous goods – as certain shipments of radioactive sources are enshrined by the regulatory framework. Another “avenue” of the awareness raising has also been the “Radiation News” – a quarterly type of the bulletin – intended primarily for domestic users of radiation sources – some of them involved also in transport-related topics. On the international parquet, the Slovene decision to join the EACA (European Association of Competent Authorities for safe transport of radioactive material) in 2015-2016 was proven to be a salient gain; the exchange of different transport-related issues, sharing good practices regionally and a pool of the group’s expertise has been a kind of “booster”.
To shortly embrace the best vision for the future, it is understandable for a small country, its regulators and their stakeholders to nurture synergies and continuous improvements. A handful of (parallel) actions have been in the pipeline – which may spur up further-on the issue of safe transport of radioactive material, having in mind sustainable approaches but also “learning by doing” and taking care of (non-unlimited) human resources and expertise in the sphere. SNSA will (also through EACA, mentioned above) collate and exchange experiences on the lessons-learned from the transport-related events. ADR (“2021 Edition”) is practically knocking at the door, envisaged to interweave numerous new requirements for dangerous goods – including radioactive material – that will need to be understood and taken into account in the future.

09.09.2020 08:30 Multinational and shared back-end approaches

Multinational and shared back-end approaches - 2001

Nuclear Power and Nuclear Technologies can Benefit from Regional Implementation of Multinational Approach

Charles McCombie, Neil Chapman, Ewoud Verhoef

ARIUS - Association for Regional and International Underground Storage, Mellingerstrasse 207, 5405 Baden, Switzerland


Nuclear energy is a proven low carbon technology that can provide the dispatchable electricity needed to stabilise national grids with increasing shares of renewables. The coming generation of small nuclear reactors (SMRs) can also make nuclear electricity available to countries without an extensive national grid. Geological disposal facilities (GDFs) play a key role in ensuring acceptance of the continued and expanded use of nuclear energy. They are also a necessity for non-nuclear power nations employing other technologies that produce small quantities of long-lived radioactive wastes. In all waste management programmes, implementing a GDF is a challenging task requiring sensitive stakeholder interactions and significant funding. For large nuclear power programmes, the costs are not the major problem since these can be covered by a minor charge on the kilowatt hours produced. For small or new programmes, the societal and economic challenges are both large. Multinational repositories (MNRs) disposing of radioactive wastes from several countries can provide a solution. MNRs can result from shared projects, from take-back of spent fuel, or from commercial initiatives.

This paper briefly reviews past and current initiatives devoted to promoting MNRs. It highlights the constructive role taken by the IAEA and the direct support given by the European Commission (EC) and by IFNEC. Comprehensive multinational studies involving over a dozen countries have been carried out over the past decades by the Arius Association and the ERDO Working Group, and several national programmes have adopted the MNR concept as part of a dual track approach in their national waste management strategy. An important organisational development is currently underway. ERDO has operated from 2009 as an informal working group – although backed and financed by countries at ministerial levels. Currently, efforts are underway to establish a formal legal entity, the ERDO Association with dedicated facilities and personnel. Multinational RWM concepts, with the considerable environmental, security and economic advantages that they offer, are becoming ever more firmly embedded within the international development landscape for nuclear power and other nuclear technologies.

09.09.2020 08:45 Multinational and shared back-end approaches

Multinational and shared back-end approaches - 2003

Deep geological repository and ideas on shared solution in Slovak national program

Adela Mrskova


The Slovak Republic had to start addressing the issue of spent nuclear fuel management intensively after the separation of the Czechoslovak Republic. During the almost 25-year history of the DGR development program in Slovakia, various ideas on national as well as international solution were considered. General history of the DGR development program in Slovakia will be presented with a brief description of the results achieved. As the development of DGR is not only a demanding multidisciplinary technical-engineering task but also by its nature as an intergenerational project, many socio-economic issues must also be taken into account when looking for a solution.
After the implementation of the EU Directive 2011/70/EURATOM the National programme for SF and RAW management in Slovakia became a leading document for the back-end of the nuclear fuel cycle in Slovakia replacing the National strategy for peaceful use of nuclear energy. Its ongoing update should reflect both previous experience and solutions to future challenges.

09.09.2020 09:00 Multinational and shared back-end approaches

Multinational and shared back-end approaches - 2005

RW Management in Croatia: Long Term Storage as Flexible Option

Alemka Knapp, Ivica Levanat, Diana Šaponja-Milutinović

Zagreb University of Applied Sciences, Vrbik 8, 10000 Zagreb, Croatia


RW Management in Croatia: Long Term Storage as Flexible Option

Zagreb University of Applied Sciences
Vrbik 8, 10 000 Zagreb, Croatia,,

Croatian plans for management of low and intermediate level waste (LILW) from Krško, presented in the recently prepared 3rd Revision of the Krško NPP Radioactive Waste and Spent Fuel Disposal Program, are based on the Croatian National Program for the Implementation of the Radioactive Waste Management (RWM) Strategy adopted in 2018. Two RWM facilities are planed:
1. A storage facility to be established by 2023, which will operate for about 40 years
2. A disposal facility which will start operation in 2051
The authors argue that better and more flexible solution would be:
1. To extend the storage facility operation to about 100 years
2. The repository establishment should be postponed accordingly
In present circumstances, when the storage facility may be needed in a couple of years from now, it is too late to call for major changes of the Croatian RWM program.
Therefore, in this paper we discuss extended storage as a “soft” modification of the RWM program, which would not have to interfere with many already planned aspects of LILW management (store location, waste packaging etc.). The only requirement on the program is to allow for the possibility of building additional storage space adjacent to the planned store, sometime around 2050, in order to postpone repository establishment.
Regarding financial aspects, it is first noted that the nominal costs for the prolonged storage scenario would be increased (extended operation, more compensations to the local community, bigger storage facility).
But discounted expenses would not increase appreciably, or may even decrease, at the discount rate of 3% that is suggested in the 3rd Revision as the internal rate of return (IRR) suitable for the Croatian program financing arrangements.
If the option of extended storage, proposed by the authors, is recognized in the Croatian RWM program, it would enhance flexibility and the robustness of the program. The paper discusses positive impacts on the feasibility of timely repository establishment, on the issue of public acceptance in the local community and other potential interactions with RWM program execution.
In order to provide for additional flexibility, different time intervals of extended storage are analyzed.

Keywords: radioactive waste, long term storage

09.09.2020 09:15 Multinational and shared back-end approaches

Multinational and shared back-end approaches - 2002

Third Revision of the Krško NPP Radioactive Waste and Spent Fuel Disposal Program Jointly Prepared by Two Expert Organisations from Slovenia and Croatia

Leon Kegel1, Zdenko Vrankić2, Sandi Viršek1, Hrvoje Prpić2, Andrea Rapić2, Goran Kukmanović2

1ARAO – Agency for Radwaste management, Celovška cesta 182, 1000 Ljubljana, Slovenia

2Fund for financing the decommissioning of the Krško Nuclear Power Plant and the disposal of NEK radioactive waste and spent nuclear fuel, Radnička cesta 47, 1000, Zagreb, Croatia


Preparation of joint Krško Nuclear Power Plant (NPP) Radioactive Waste and Spent Nuclear Fuel Disposal Program (Disposal Program) and joint Krško NPP Decommissioning Program and regular revisions of both programs is stipulated in the Agreement between the Government of the Republic of Croatia and the Government of the Republic of Slovenia on the Regulation of the Status and Other Legal Relations Regarding the Investment, Exploitation and Decommissioning of the Krško NPP (Bilateral Agreement). Intergovernmental Commission (IC) monitors implementation of the Bilateral Agreement, and is responsible for starting, supervising and final acceptance of programs and its revisions.
Third revision of Disposal Program was prepared in accordance with IC approved Terms of Reference (ToR) by expert organizations: Fund for financing the decommissioning of the Krško NPP in Croatia and ARAO in Slovenia in cooperation with NEK. Basis for the Third revision were new operational and decommissioning inventory assessments, and new circumstances that developed since the last revision (updated preliminary decommissioning plan, new national strategies and programs, extension of Krško NPP’s lifetime, on-site SF dry storage, project development and knowledge in the area of RW and SF management/disposal, possible division of RW and other). Main objective was to construct technologically feasible disposal scenarios and to assess nominal costs of RW and SF management / disposal. Main boundary conditions for Third revision of Disposal Program were:
- joint strategy for SF management including dry storage on site and joint deep geological repository or participation in the multinational disposal solution and
- separate solutions for management / disposal of low and intermediate radioactive waste according to the national strategies and programs.
IC nominated project Implementation Coordination Committee (ICC) for support of Third revision project and preparation of proposals for possible joint radioactive waste disposal.
In the article newly assessed inventory will be presented together with developed RW and SF management scenarios, proposal for RW division, estimated nominal costs for decommissioning, investment and operational costs of the storage and disposal facilities with contingencies and compensations, conclusions and recommendations of the Third revision.

09.09.2020 09:30 Multinational and shared back-end approaches

Multinational and shared back-end approaches - 2006

EURAD programme: effective and efficient public RD&D funding in Europe

Nadja Železnik1, Leon Cizelj2, Leon Kegel3

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

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

3ARAO – Agency for Radwaste management, Celovška cesta 182, 1000 Ljubljana, Slovenia


In 2019 the new very ambitious European Joint Programme on Radioactive Waste Management (EURAD) has started as part of the EURATOM activities under Horizon 2020 with a vision to assure “a step change in European collaboration towards safe radioactive waste management (RWM), including disposal, through the development of a robust and sustained science, technology and knowledge management programme that supports timely implementation of RWM activities and serves to foster mutual understanding and trust between participants.”
EURAD programme supports the implementation of the Waste Directive in EU Member-States, taking into account the various stages of advancement of national programmes. National RWM programmes across Europe cover a broad spectrum of stages of development and level of advancement, particularly with respect to their plans and national policy towards implementing geological disposal. Programmes differ significantly depending on the national waste inventory, with some member states only responsible for relatively small volumes of medical and research reactor wastes, compared to others that have comparatively large and /or complex waste inventories from large nuclear power (and fuel reprocessing) and defence programmes. Programmes also differ significantly in the way in which they are managed, particularly with respect to the national policy and socio-political landscape with respect to longer-term storage and geological disposal.
The paper will present the first phase of EURAD programme which will last until 2024 with work packages and their objectives, organizations of work and governance, and the first results.

09.09.2020 09:45 Multinational and shared back-end approaches

Multinational and shared back-end approaches - 2004

The Relevance to the Central Europe Region of Multinational Approaches to Addressing the Challenges Presented by the Back End of the Fuel Cycle

Tomaž Žagar1, Sean Tyson2, Robert Mussler3

1Društvo jedrskih strokovnjakov Slovenije, Jamova 39, 1001 LJUBLJANA, Slovenia

2U.S. Department of Energy, 1000 Independence Ave., SW, DC 20585 Washington, USA-Washington D.C.

3Secretariat of Governmental Committee for Nuclear Emergency Preparedness, Magyorodi ut.43, 1149 BUDAPEST, Hungary


(1) Tomaž Žagar, Head of Planning and Control, GEN energija d.o.o., Slovenia, IFNEC RNFSWG Co-Chair ( corresponding author
(2) Sean Tyson, Office of International Nuclear Energy, Office of Nuclear Energy, U.S. DOE, USA, IFNEC RNFSWG Co-Chair (
(3) Robert Mussler, Nuclear waste policy consultant, IFNEC Technical Secretariat (

Like the power generation options of fossil fuels (oil and gas), nuclear power can provide reliable, 24/7 generation. Unlike fossil fuels however, the wastes produced are completely contained and managed with no detrimental releases to the atmosphere. Also unlike fossil fuels, in comparison the wastes produced per unit of power generation, wastes from nuclear are very small in volume and easily contained and managed.
The trade-off for having much smaller volumes and being more easily contained and managed is the resulting High-level Radioactive Waste (HLW) and/or Spent Fuel (SF) that requires technologically advanced management procedures and adequate isolation to prevent any negative impact on the environment and/or human health. These management procedures and isolation measures are effective, well developed and come with significant experience.
A key element in the management of wastes from nuclear power is the disposal of SF in a Deep Geological Repository (DGR). The DGR is internationally recognised as the most technologically developed and safest approach for the final step in HLW and/or SF management. However, the development of a DGR involves high fixed costs that carry an associated economy of scale. This means that smaller nuclear programs generating small amounts of waste could benefit through sharing solutions or other multinational approaches to the challenges presented by the back end of the fuel cycle. Such multinational approaches are especially interesting in certain geographical regions of the world, where several individual smaller nuclear programs are close together.
International Framework for Nuclear Energy Cooperation’s (IFNEC) Reliable Nuclear Fuel Services Working Group (RNFSWG) is one of the international forums working on developing an understanding of various multinational approaches to addressing the challenges presented by the back end. This work is of great relevance to the larger Central European region and relevant to the target audience of this conference.