NENE conference 2021

Invited lectures - 101

Building the Future on Solid Foundation - The Value of Multigenerational Nuclear Community

European Nuclear Society, Avenue des Arts 56, B-1040 BRUXELLES, Belgium

In the present moment, nuclear science and industry reunites at least 4 generations of employees. At some point, each one of them felt or will feel unique and outstanding, not comparable with any other. The youth dreams of first nuclear professionals revolving around dense, potent source of energy capable to provide electricity to each home in increasingly affordable and innovative way, gets a new dimension for the current generation of young professionals: a potential of halting the climate change in a way that is respectful for environment and societies. Fulfilling this potential is what the current novices will be facing all their careers.
Ensuring the central role of nuclear energy in the face of the climate and developmental challenge takes much more than vision and charisma of the youth. It also requires skillful application of experience and lessons drawn from successes and failures. On the other hand, this vision, its scope and timescale goes beyond one career or even one lifetime – it will need to be passed to those who come after.
Nuclear, with all its initial and current challenges and opportunities, needs complex approach to building a well-functioning, multigenerational community that results not only in cooperating and tolerating one another. The inclusion, adapted approach and drawing the best traits of each of the generations, starting with those who remember the first Magnox started up to those who soon will hear about neutrons for the first time, will facilitate answering to the issues of the current times while ensuring smoothness, resilience and sustainability of nuclear science and industry.

06.09.2021 17:10 Invited Abdesselam Abdelouas

Invited lectures - 102

SAfe and REliable Nuclear Applications (SARENA)

Abdesselam Abdelouas

École des Mines de Nantes, 4, rue Alfred Kastler - La Chantrerie, CS 20722 44307 Nantes cedex 3, France

abdeloua@subatech.in2p3.fr

The Erasmus Mundus joint Master of Science in SAfe and REliable Nuclear Applications (SARENA) is a European cooperation and mobility program in the field of higher education. Unique in Europe, the Master SARENA aims to develop scientific, technical and management skills enabling engineers to work in all nuclear energy-related fields and applications.
Initiated and coordinated by the French Engineering school IMT Atlantique, it is supported by a consortium of four European universities recognized for the excellence of their research and training offer in the field of Nuclear Science and Engineering, namely Universidad Politécnica of Madrid (Spain), University of Technology Lappeenranta (Finland) and University of Ljubljana (Slovenia). Responding to the needs expressed by the actors in the Nuclear sector, SARENA also enjoys broad support from industries (EDF, Orano, ASSYSTEM, GEN Enerjia), public institutions and research organizations (Andra and CEA especially for France).
Fully taught in English, the 2 years Master includes three complementary semesters of specialization followed by a semester dedicated to the thesis project in industry or within a research laboratory. The Master offers two tracks and therefore two different mobility paths: 1. Radioactive Waste Management and Decommissioning track with a mobility at IMT Atlantique and Universidad Politécnica of Madrid. 2. Nuclear Reactors Operation and Safety tarck with a mobility at IMT Atlantique, University of Technology Lappeenranta and University of Ljubljana. At the end of the Master, students are awarded a double degree.
Finally, having received the Erasmus Mundus label for 4 intakes, the Master SARENA benefits from 66 excellence scholarships covering all participation, installation and living costs of the students.

07.09.2021 08:30 Invited Elia Merzari

Invited lectures - 104

Full core Computational Fluid Dynamics: Recent advancements and Demonstrations

Elia Merzari

Pennsylvania State University, Department of Nuclear Engineering, 228 Hallowell Building, University Park, PA 16802-4400, PA 16802-4400, USA-Pensylvania

ebm5351@psu.edu

GPU-based supercomputing is enabling a significant advancement in CFD capabilities for nuclear reactors. In fact, pre-exascale GPU-based super-computers such as ORNL’s Summit are allowing for the first time to perform full core CFD calculations with URANS and LES approaches.

However, computing power alone is however only part of the story. Several advances in computational fluid dynamics methods and their implementation had to be introduced to achieve this milestone. As part of this talk we discuss in particular the development of NekRS, a novel GPU-oriented variant of Nek5000, an open source spectral element code in development at Argonne National Laboratory. NekRS delivers peak performance for key kernels on the GPUs, and good scaling performance even on GPU architectures. Recent performance measurements showed that NekRS, when running on GPUs, outperforms the CPUs by 40x.

The combination of extreme computational power and novel algorithms has enables a significant leap forward. In this talk we examine in particular two recent demonstrations : (i) the full core URANS simulation of the Nuscale Small and Modular reactor (SMR) including the modeling of spacer grids, (ii) the full core simulation of the active region of the pebble bed core of the Mark-I Berkeley’s Flouride-Cooled High Temperature (FHR) reactor. We also examine how these calculations are being used to improve the fidelity of more traditional approaches such as porous media models for pebble beds.

In the final part of this talk we discuss the future of the field ,as exascale systems are due to come online in the near future.

07.09.2021 14:20 Invited Alireza Haghighat

Invited lectures - 103

A New Paradigm for High-Fidelity 3-D Reactor Kinetics Simulations

Alireza Haghighat1, Valerio Mascolino1, 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

haghighat@vt.edu

There is a significant need for 3-D steady-state and transient neutron transport algorithms
and codes that yield accurate, high-fidelity solutions using reasonable computing resources
and time. These tools are essential for modeling innovative nuclear systems, such as next generation reactor designs, as they allow fast but accurate simulation of these systems in a large variety of configurations. The existing methods generally compromise heavily between
accuracy and affordability in terms of computation times.

We have developed a novel time-dependent algorithm based on the Transient Fission Matrix (TFM) method and implemented into the RAPID code system. These new algorithms have been computationally verified using several computational benchmarks and experimentally validated using the JSI TRIGA Mark-II reactor.

We have demonstrated that RAPID can yield high-fidelity time-dependent solutions in seconds and minutes on a single computer core by pre-calculating a database of response functions and coefficients, while such calculations are impractical using standard Monte Carlo or deterministic methods.

08.09.2021 08:30 Invited Uwe Stoll

Invited lectures - 105

Current Challenges in Reactor Safety Research

Uwe Stoll

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

uwe.stoll@grs.de

The nuclear sector is currently facing challenges from two different directions. On the one hand, safe operation must continue to be guaranteed for systems with a service life of several decades. On the other hand, new reactor concepts are being developed, for whose novel systems the necessary reliability must first be proven. Germany's central technical support organization GRS investigates the various issues that arise from these challenges. Typical research topics include the behavior of passive safety systems in new reactor concepts as well as the investigation of vibration phenomena in BWR reactor cooling circuits and the enhanced fuel rod cladding corrosion in PWR.

09.09.2021 08:30 Invited Bernard Bigot

Invited lectures - 106

The ITER Project: moving fusion power closer to reality

Bernard Bigot

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

bernard.bigot@iter.org

The ITER project is a collaboration of 35 countries to build the world’s largest fusion energy device, to demonstrate the feasibility of fusion power at an industrial scale. Recent years have seen rapid progress in construction, manufacturing, and – starting in mid-2020 – assembly phase. Currently, about 73% of the overall work required to achieve First Plasma has been completed. On the ITER worksite, this progress is visible firstly in the completion of key buildings and infrastructures. The tokamak building was declared ready for equipment as of Spring 2020. Commissioning of the connection to the EU grid and the substation for the steady-state electric network is complete; and commissioning of other key plant support systems (e.g., cooling water, cryoplant, pulsed power for magnet systems) is well underway.

On the manufacturing front, progress is equally impressive. The base and lower cylinder of the cryostat have been installed and welded together in the tokamak pit; the upper cylinder is also complete and in storage. The first two poloidal field (PF) coils were turned over to ITER in early 2020; PF6 has been installed, PF5 will be installed in the coming weeks, and four additional PF coils are in advanced stages. Factory tests of the first central solenoid (CS) module will be completed; the first two modules will be shipped in the coming months, and five more modules in various stages of manufacturing. The first 7 of 18 toroidal field (TF) coils have been delivered to ITER, starting in April 2020, and progress on the remaining TF coils is ongoing. The thermal shield sections have been delivered, and the lower cryostat thermal shield has been installed. The first vacuum vessel sector arrived onsite in August 2020, and – together with two TF coils and a section of the vacuum vessel thermal shield – is being incorporated into the first sub-assembly. An important aspect of this progress is that, for most of the major first-of-a-kind components, the capability has now been demonstrated to fabricate according to ITER’s demanding specifications.

With this progress achieved, ITER is well into Assembly Phase, and the next few years will be dominated by assembly, installation, system commissioning and integration. Following First Plasma, the path toward the achievement of the Q = 10 project goal has been consolidated in a Staged Approach, with all systems to be installed before the start of full fusion power operation in 2035. While the Covid-19 pandemic has had impacts – both on the ITER worksite and in factories in Member countries – the project remains largely on track; even if some delays occur with regard to the goal of First Plasma in 2025, we expect to stay fully on track for Full Fusion Power in 2035.

Highlights from each of these areas (manufacturing, commissioning, tokamak assembly) will be presented along with the updated status and plans.

06.09.2021 18:10 Advances in nuclear technology

Advances in nuclear technology - 201

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

Patrick Fragman1

1Westinghouse Electric Company , 1000 Westinghouse Drive, PA 16066 Cranberry Twp, USA

The challenge set by utilities and regulators for an improved Generation III+ class of reactors was to achieve increased levels of safety, constructability and operability, while reversing the trend for greater complexity in reactor design. The AP1000 pressurized water reactor sets a new standard for nuclear power plant simplicity and safety using all passive safety systems, requiring no AC power or operator actions to maintain plant cooling for at least 72 hours after a postulated accident event. The AP1000 plant also utilises advanced, modular construction methods, further optimised for the efficient delivery of all future projects. The AP1000 plant is now fully proven with four units in commercial operation after exceptional first-of-a-kind testing and commissioning programs, first cycle performance and record-breaking first refuelling outages. Two more units are expected to reach commercial operation in late 2021 and 2022. With the experience of the first wave of projects incorporated, the AP1000 plant is the proven, advanced generation III+ technology option for the Slovenian new build program and beyond.

06.09.2021 18:30 Advances in nuclear technology

Advances in nuclear technology - 212

Evolution of VVER reactors: from prototype to 3+ generation. Technological development

Alexander Renev

ROSATOM State Atomic Energy Corporation , Bolshaya ordynka 24/26, 119017 MOSCOW, Russian Federation

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. Now Rosatom has 35 units in 12 countries in its overseas projects implementation portfolio. For the past 15 years, Rosatom has commissioned 17 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. VVER-TOI design is an evolution of the VVER-1200 one with higher capacity. The first VVER-TOI is under construction in Russia, Kursk. The “VVER-TOI” Design has been developed to integrate the optimal information solutions and engineering approaches, elaborated of late by JSC “Atomenergoproekt” experts. The integrated power unit data management system, developed within the Design framework, is in fact one-off solution with no analogs worldwide.

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.

06.09.2021 18:50 Advances in nuclear technology

Advances in nuclear technology - 202

SMR Safety – Advantages and Challenges

Gérard Cognet, Jan Bartak, Gianni Bruna

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

Looking at the historical evolution of nuclear power reactors, the long-term trend so far has been to increase the size of the reactors in order to maintain their economic competitiveness through economies of scale. However, over the last five years, advanced reactor concepts, among which SMRs, have progressed faster than anyone predicted ten years ago, and it is highly likely that over the next ten years we will see the construction and operation of several FOAK (First of a kind) SMRs and the development of a global supply chain to support them. One of the reasons for this keen interest is that SMRs could become one of the main drivers of the deep decarbonisation of global economy, thanks to their versatility, flexibility, and ability for cogeneration (electricity plus heat and/or hydrogen). This interest and the diversity of applications (electricity in remote areas, low and high enthalpy heat for heating and industry, water desalination, hydrogen production) explain that there are more than 70 different design concepts under development around the world with different technology and licensing readiness levels. Moreover, most SMR designs rely on higher levels of intrinsic safety and/or passive safety systems compared to Gen III/III+ LWRs which should facilitate their acceptability by the public and allow their operation within existing industrial sites or closer to large cities.
However, it is important to underline that most of the SMRs still must overcome significant technical hurdles in domains such as nuclear fuel reliability, materials behaviour, component manufacturing and more globally safety assessment and licensing. In this regard, it can be noted that, except the Russian barge-mounted SMR, the first of which (Akademik Lomonosov) is already operational, all the other SMR designs have still to be licensed in a context where the existing regulatory frameworks have to be reviewed and, presumably modified to some extent, to make them applicable to this new type of reactors.
In this context, this paper aims at presenting a general overview and an assessment of safety features of different families of SMRs (light water cooled, high temperature, fast neutron and molten salt) without bias and independently of any of the designers.
Although it is extremely difficult to compare the large and quite diverse family of SMRs with conventional and advanced large reactors (even more so from a security point of view), it is nevertheless possible to identify the conditions and opportunities for safe development and further improvements in the design and operation of SMRs, taking advantage of their inherent safety features, low power and small size. To achieve this goal, this document will be organized as follows:
• The first section will provide an overview of the main general design and operational features shared by most SMRs;
• The second section will:
o remind and summarize the general safety principles that any SMR must comply with;
o address the safety features which are, to a certain extent, shared by the majority of SMRs;
o underline several still open safety issues for SMRs;
• The third section will delve more deeply into the safety features (presented as advantages vs drawbacks) for four selected SMR types: light water, high temperature, molten salt and fast neutrons.
In conclusion, the main items addressed in the discussion will be summarized.

06.09.2021 19:10 Advances in nuclear technology

Advances in nuclear technology - 203

Teplator DEMO nuclear heating for Prague central heating network.

Radek Skoda, David Masata, Tomas Peltan

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

A case study focusing on the Teplator DEMO nuclear heating deployment for the existing Prague district heating system is presented. The case study aims at existing gas/waste/coal district heating Prague portfolio as well as at the network and existing nuclear licensed sites which are off the city center and not far from existing heat pipelines. A broad technical and cost study is studied to assess feasibility of nuclear heating source for existing conditions.

07.09.2021 09:10 Thermal-hydraulics & CFD I

Thermal-hydraulics & CFD - 601

Classification and Resolution Adaptive Drag Modelling of Gas-Liquid Interfaces with a Multifield Two-Fluid Model

Richard Meller1, Matej Tekavčič2, Benjamin Krull1, Fabian Schlegel1

1Institute of Fluid Dynamics, Helmholtz-Zentrum Dresden - Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden, Germany

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

r.meller@hzdr.de

Reliable techniques for the numerical simulation of gas-liquid flows at industrial scales are of great interest for safety analysis and efficiency optimisation, e.g. in nuclear power or metal processing industries. This type of simulation is hard to carry out due to the immense range of scales, which is spanned by interfacial and turbulent structures. For this purpose, hybrid morphology-adaptive numerical frameworks are being developed in the recent years, combining different well established numerical methods for individual flow regimes. The present work follows the approach of Meller et al. (Int J Numer Meth Fluids. 2021; 93: 748-773), who utilise the Euler-Euler approach to statistically describe multiphase structures in dispersed flow regimes, such as bubbly flows, as a basis. At the same time, regimes with resolved gas-liquid interfaces, such as large rising gas bubbles or horizontal interfaces in stratified flows, are captured by means of a Volume-of-Fluid-like method (Tekavčič et al., Nucl Eng Des. 2021; 379: 111223). A fully morphology-adaptive numerical framework needs to comprise transitions between the aforementioned regimes. Hence, the limits of both underlying basic numerical approaches need to be pushed towards and beyond an overlapping region of grid resolution with adequate predictive power, such that the whole spectrum of length scales is covered, forming the basis of morphology transitions.
For this purpose, the present work focuses on the extension of the Volume-of-Fluid methodology towards reliable resolved simulations of gas-liquid interfaces with very coarse grid resolutions. By means of the underlying two-fluid model, an interfacial slip velocity in the interface region becomes generally possible and can be chosen physically motivated. The flow in the vicinity of the interface needs to be classified. For this purpose, the latter is categorised to be of shear type, stagnant type or in between the two. Furthermore, a dimensionless grid spacing is evaluated based on the shear stress across the interface, similarly to the y+ value for the cell thickness of wall-bounded flows. Besides that, interface curvature is considered in relation to grid spacing. From these information, a degree of under-resolution of the interface is determined, which subsequently serves as a criterion for the drag modelling framework. On this basis, interfacial drag coupling is manipulated, such that interfacial slip can take place in the direction tangential to the interface, whenever required. While the interfacial drag formulation of Štrubelj and Tiselj (Int J Numer Methods Engng. 2011; 85: 575-590) is used in case of proper resolution, the closure formulations of Porombka and Höhne (Chem Eng Sci. 2015; 134: 348-459) or Marschall (Technical University of Munich, PhD Thesis, 2011) are considered for portions of the computational domain, where interfaces are classified to be under-resolved. The functionality of the described procedure is validated in cases of 2D and 3D rising gas bubbles, considering their shape and rising velocity. Moreover, gas and liquid velocity profiles of a stratified flow serve as a validation in an additional flow regime.
In this way, the numerical prediction of the gas-liquid interface is improved, pushing the limit of spatial resolutions with adequately reliable predictions towards extremely coarse computational grids, which is the prerequisite for efficient numerical simulations in large-scale applications.

07.09.2021 09:30 Thermal-hydraulics & CFD I

Thermal-hydraulics & CFD - 602

GPU-based accelerated computation of coalescence and breakup frequencies for polydisperse bubbly flows

Gasper Petelin1, Jeffrey Kelling2, Ronald Lehnigk3, Gregor Papa1, Fabian Schlegel3

1Jožef Stefan Institute, Computer Systems Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

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

3Institute of Fluid Dynamics, Helmholtz-Zentrum Dresden - Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden, Germany

gasper.petelin@ijs.si

Polydisperse bubbly flows appear in many industrial applications and are particularly relevant for nuclear and process engineering. A popular approach for their simulation is the two-fluid model, which requires information about the bubble size. An associated challenge is to incorporate the effects of coalescence and breakup on the size distribution, which can be tracked by a population balance equation. It can be solved by the method of classes, which splits the bubble population into a finite number of size groups. However, this technique is expensive because the source term assembly involves the computation of coalescence and breakup frequencies between all bubble size pairs. This work focuses on improving the performance of the class method implementation within the OpenFOAM Foundation release (www.openfoam.org), an open source computational fluid dynamics software that allows for domain decomposition and parallel execution on multiple central processing units (CPUs). Here, the parallel computation of the coalescence and breakup frequencies is shifted to graphics processing units (GPUs) using the Nvidia CUDA framework. Calculations are done asynchronously with overlapping data transfers, treating size pairs as independent units of work that are computed in parallel. The coalescence and breakup frequency computation on GPUs leads to a significant speedup compared to the pre-existing implementation. The improvement is demonstrated for a co-current two-phase flow in a vertical pipe.

07.09.2021 09:50 Thermal-hydraulics & CFD I

Thermal-hydraulics & CFD - 603

Effect of heat flux on axial distribution of vapour volume fraction and bubble sizes at flow boiling in horizontal annulus

Boštjan Zajec, Boštjan Končar, Leon Cizelj

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

bostjan.zajec@ijs.si

Subcooled flow boiling will be investigated in a horizontally placed 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 annular test section is constructed as a double-pipe heat exchanger, where the heat is transferred from the heating water flowing inside the inner copper pipe to the working fluid flowing through the annular gap between the inner pipe and outer transparent pipe that enables visualization of the boiling flow. Refrigerant R245fa is used as a working fluid as it boils at lower temperatures and lower pressures than water.
The main objective of this study is to investigate the effect of heat flux variation on the boiling flow patterns at approximately constant inlet flow conditions of the working fluid (fixed mass flux and inlet fluid temperature). The variation of heat flux will be controlled by changing the mass flux of the heating water flow. Subcooled flow boiling of R245fa will be recorded by a high-speed camera and images of the boiling flow patterns will be then analysed using image processing to determine bubble size distributions and the resulting distributions of vapor volume per bubble size. The axial distributions of vapor volume fraction and bubble sizes will be determined by shifting the camera position along the test section. The experimental setup, analysis methods and measurement results will be presented and discussed.

07.09.2021 11:00 Severe accidents I

Severe accidents - 402

KIT-JIMEC experiments to investigate jet impingement on a core catcher bottom and ablation process

Walter Tromm, Xiaoyang Gaus-Liu

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

walter.tromm@kit.edu

In 2019, the research activities of KIT in the field of experimental severe accident research on fast reactors were concentrated on the European ESFR-SMART project. Within this project two large-scale JIMEC experiments have been performed to investigate the thermal ablation kinetics of an internal core catcher material in a SFR reactor. Besides this, the planning work for LIVE-ESFR tests to study the interaction between the corium simulant and the sacrificial simulant of the core catcher started. It has been decided to construct and build a new test vessel with down-scaled geometries similar to SFR core catcher design.
The actual safety design of a Sodium Fast Reactor (SFR) in the case of a postulated severe accident includes to remove the corium from the core by corium transfer tubes and to collect it in the lower head in an in-vessel core catcher. It is assumed that at first a metallic corium melt jet would impinge on the core catcher surface and could ablate the core catcher material. Experimental data is needed to simulate this ablation behaviour of a long duration melt jet impinging the core catcher material. A particular behaviour can be studied when a molten pool is created (“pool effect”) at the impact point that could reduce the heat transfer at the jet – material interface. This phenomenon has not yet been studied suffienctly. Therefore, the ITES-SAR (Institute for Thermal Energy Technologies and Safety - Severe Accident Research Group) team has adapted the existing MOCKA test facility to perform two JIMEC (Jet Impingement on Metallic Core Catcher) experiments in the frame of the European ESFR-SMART project. JIMEC-1 and JIMEC-2 tests investigate the characteristics of the interaction of a metallic melt jet with the core catcher bottom plate material in a SFR reactor design with prototypical material. The objectives of the experiments are to deliver experimental data on the interaction of melt jet parameter and erosion dynamics. The melt jet parameters were jet temperature, jet velocity and jet diameter. The erosion dynamics in the core catcher bottom is the erosion velocity and the timing of pool effect. The experimental results will be used for developing new correlations which could be used in codes for simulation of the ablation kinetics for SFR core catcher concepts.
The two JIMEC experiments which have been performed in summer 2019 in the adapted MOCKA test facility, will be described in detail in this paper and conclusions will be drawn from the results on possible core catcher concepts.

07.09.2021 11:20 Severe accidents I

Severe accidents - 403

Long-term containment cooling in Fukushima unit 1: insights into consequences from sensitivity studies

Luis E. Herranz, Rafael Bocanegra

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

rafael.bocanegra@ciemat.es

The Fukushima accidents have already provided and will continue providing relevant insights into consequences of major management actions conducted at the time. Based on the analyses carried out within the OECD/BSAF2 project with MELCOR 2.2 on Unit 1, this study explores what the alternative water injection, after roughly 10 days after the accident onset, would have meant in case it had occurred earlier or later in the sequence, or it had been stronger or weaker than it supposedly was. Containment pressure and fission product release to the environment are the main figures of merit in this study.

07.09.2021 11:40 Severe accidents I

Severe accidents - 404

Experimental Investigation on the Post-Dryout Behaviour of Homogenous Debris Beds

Markus Petroff, Rudi Kulenovic, Joerg Starflinger

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

markus.petroff@ike.uni-stuttgart.de

Severe accidents of light water reactors with a loss of coolant can result in an overheating of the fuel rods, the loss of core integrity and the core relocation into the lower plenum of the RPV. Due to the possible presence of residual water a debris bed can be formed by fragmentation. The removal of decay heat of the debris bed is of prime importance to prevent any damage to the RPV. The DEBRIS test facility at IKE, University of Stuttgart, has been modified to investigate the post-dryout behaviour of volumetrically heated cylindrical debris beds (200 mm inner diameter, 700 mm height).
Former experimental investigations indicated the presence of quasi-steady states in the post-dryout boiling regime where no further temperature escalation was observed. Experimental investigations of such quasi-steady state conditions as well as the accompanied occurring processes in the post-dryout regime are only insufficiently available. Therefore, especially for the model validation of COCOMO-3D implemented in the thermal-hydraulic system code ATHLET, a specific extension to the existing experimental database is required.
The paper gives a brief outline on the state-of-the-art of previous experimental research related to dryout experiments in the post-dryout regime and describes in detail the modified DEBRIS test facility. First experimental results of the post-dryout behaviour of a monodispersed particle bed are presented and discussed within this paper.

07.09.2021 12:00 Nuclear power plant operation

Nuclear power plant operation - 507

Long term operation of NPP Krško and related challenges

Stanko Manojlović

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

stanko.manojlovic@nek.si

NEK has prepared a comprehensive program for ensuring safe long term operation of the plant beyond a time-frame established in the license conditions, design limits, safety standards. Additionally, overall plant responsibilities for the implementation of Long Term Operation (LTO) Program as a living activity have been defined.
In general, NEK LTO Program consists of various NEK programs, procedures and activities which ensure that all intended functions of Systems, Structures and Components (SSC) are in place, that all SSCs managed for long term operation are recognized, properly reviewed and managed in such a way, that they will fulfill their intended function until the end of NEK entire operating lifetime.
Preconditions for long term safe and stable operation require planning and efficient implementation of personnel processes, timely recruitment and systematic development of all employees. Knowledge and experience must be managed effectively and comprehensively. There is high awareness that only professionally qualified and competent individuals are a prerequisite for safe, efficient and high-quality implementation of various work processes which ensure overall progress in all areas of work.
Multiple actions have also been taken in the past decade to allow safe plant operation, such as continuous maintenance on SSCs, multiple modifications to replace equipment and to improve performance, implementation of the Safety Upgrade Program (SUP), completion of the Periodic safety review and other actions which improve NEK safety and reliability.
Three key challenges are in progress before 2023, which are: the third periodic safety review, environmental impact assessment and pre-SALTO mission, which will assess the safety aspects of long term operation.

07.09.2021 12:20 Nuclear power plant operation

Nuclear power plant operation - 501

EXPERIENCES IN SYSTEMATIC APPROACH TO PROCEDURES VALIDATION IN SIMULATOR

Nicolas Moyano, Alejandro Arroyo, Borja Hervas

Tecnatom, s.a., Avda. Montes de Oca, 1, 28709 San Sebastian de Los, Reyes, Madrid, Spain

nmoyano@tecnatom.es

Nuclear Power Plant Krško (NEK) decided to take steps to upgrade safety measures to prevent severe accidents and to improve the means to successfully mitigate their consequences. The content of the program for NEK Safety Upgrade is consistent with the nuclear industry response to the Fukushima accident, addressing Design Extended Conditions (DEC) events. As part of this program, Tecnatom was involved in several activities related with the engineering, design, procurement, and operation of the new Emergency Control Room (ECR). These activities were supported by Tecnatom knowledge of plant operation and human factors engineering.
In the same way, Tecnatom has recently provided an independent review of NEK Emergency Operating Procedures (EOPs) revision, which includes the new plant DEC systems and equipment. Both contributions were culminated with the execution of an Integrated System Validation (ISV) at Krško Full Scope Simulator (KFSS). The ISV is the final stage of the systematic approach followed by Tecnatom to enhance the ECR design and NEK procedures. Prior to the it, partial validation sessions using mock-ups at different design stages were conducted.
The systematic approach followed is in accordance with the U.S. Standard for Human Factor Engineering Program, NUREG-0711. The initial stage consists on an analysis of different subjects related to the plant modification. Existing system functions affected by the plant modifications are identified, as well as new functions. In the next step, the tasks required to fulfill those functions are defined and evaluated in terms of requirements. Finally, to conclude the analysis stage, the control room staffing is set based on tasks analysis. The main result of this stage is an Operating Sequence Diagram, a graphic representation of the task sequence to be performed during a specific scenario, including the estimated task timeline for each operator involved. The OSDs showed to be a powerful tool to unveil weaknesses in the design and procedures during the validation sessions.
The systematic approach ends with the ISV, which is defined as an evaluation using performance-based tests to determine whether an integrated system design (i.e., hardware, software, procedures, and personnel elements) meets performance requirements and supports the plant’s safe operation. A total of three ISVs were conducted: two of them related to the event of main control room evacuation and a third one related to DEC events.
All the experiences were positive, highlighting the enhancement of safety, the acceptance of the final user and the high adaptability to each project’s nature. The integration of the process within the design allowed its adjustment based on the results of the validation, which meant a valuable improvement of operations safety and human performance of the end-user.

07.09.2021 12:40 Nuclear power plant operation

Nuclear power plant operation - 502

Reliability of Power System in Slovenia – Comparison of Scenarios

Marko Čepin1, Aljaz Spelko2, Bruno Glaser3

1Faculty of Electrical Engineering, Tržaška cesta 25, 1000 Ljubljana, Slovenia

2University of Ljubljana, Faculty of Electrical Engineering, Tržaška 25, 1000 Ljubljana, Slovenia

3GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia

marko.cepin@fe.uni-lj.si

Due to the complexity of the power system, its reliability cannot be assessed by one method and cannot be represented by one parameter. Several methods exist and each can contribute to the power system reliability from its own viewpoint. Power system reliability is generally represented by adequacy and security. Adequacy deals with the power system as a static system of its components, while security deals with evaluation of power system dynamics, its transients and its stability. This paper mostly deals with the adequacy and the loss of load expectation is used together with its upgrade with the recursive algorithm and with its upgrade related to the time-dependent modelling and analysis. The loss of load expectation is a method that gives us an estimate for the power system that is important for system planning and which gives a time interval in which the power generation cannot cover consumption in a probabilistic way in a certain period. The results specify the number of hours in a year when the power generation cannot cover consumption: the smaller the loss of load expectation, the better the power system reliability.
Reliability analysis of power system in Slovenia is performed through the application of loss of load expectation for several the most applicable scenarios, namely the current configuration scenario, the scenario with new nuclear power plant JEK2, the scenario after the shutdown of the nuclear power plant Krško, the scenarios with renewable sources of energy which replace some thermal power and the mixed scenarios, where renewable sources are combined mostly with nuclear power. The scenarios with renewable sources and the mixed scenarios with renewable sources and nuclear power are further divided into several subgroups of scenarios. The periods, which are considered and analysed, include the years 2018, 2019, 2020, 2028, 2030 and 2040. The recent years are selected for analysis of the power system current state. The years 2028, 2030, 2040 are selected, because they are important milestones in strategic documents in Slovenia.
The results show that the current power system in Slovenia is reliable, with relatively small average loss of load expectation. The winter months show notably reduced reliability, partly due to a larger electric energy consumption and partly due to a smaller power generation from solar power plants, which in winter months operate with significantly smaller power compared to summer months. The reliability of the Slovenian power system improves significantly after the inclusion of JEK2. In the future scenarios, the reliability of power system in Slovenia decreases significantly without nuclear power plants. The results show that the reliability of power system cannot be achieved only by volatile and weather dependent renewable power sources of energy, but one needs to build other power plants with heavy rotation masses, e.g. nuclear power plant, which largely improves the reliability of power system.

07.09.2021 15:00 Reactor physics & research reactors I

Reactor physics & research reactors - 307

Thermal Scattering Law Data for Zirconium Hydride from First-Principles

Ingrid Svajger1, N. C. Fleming2, B. Laramee2, Gilles Noguere3, A. I. Hawari2, Luka Snoj1, A. Trkov4

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

2North Carolina State University, 2500 Stinson Dr. Raleigh, 27695-7909 NC, USA-North Carolina

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

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

ingrid.vavtar@ijs.si

Many transition metals react with hydrogen to form stable metal hydrides. Metal hydrides are technologically attractive materials due to their ability to store high density of hydrogen. A considerable interest in zirconium hydrides arises in the nuclear industry, where it is essentially used as a neutron moderator in nuclear reactors. TRIGA reactors feature unique fuel composition, i.e. a homogeneous mixture of 20 % enriched uranium and zirconium hydride (ZrH ratio near 1.6). This is a primary reason for the prompt negative temperature reactivity coefficient. Since hydrogen in the zirconium hydride serves as a moderator, most moderation occurs in the fuel element itself and only a small part in the water surrounding the fuel elements. Therefore, any change in power and in fuel temperature immediately reflect on the moderator in the fuel element. In this case, the fuel and the moderator promptly affect the reactivity of the core. Consequently, this emphasizes the need to understand the fundamental atomistic properties of zirconium hydrides.
ZrHx may exist in multiple phases with varying stoichiometry, the most significant of which are the ? phase (at room temperature dominant for 1.56 < x < 1.64) and the ? phase (at room temperature dominant for x > 1.74). Current ENDF/B-VIII.0 ZrHx nuclear data evaluations do not distinguish between phases. In the present work, we adopt first-principles calculations to obtain the thermal neutron sattering law data (TSL) of ZrH1.5 in ? phase and ZrH2 in ? phase. Using the density functional theory (DFT) capable computer code VASP, the system is modelled and relaxed to its ground state. Atomic positions are then perturbed, and the interatomic force constants are calculated. Once the force constants are obtained, they are transferred to the PHONON or Phonopy codes, which perform lattice dynamics calculations in which solutions to the dynamical matrix problem are sought. The solutions constitute the dispersion relations of the system, from which the atomic vibrational density of states (DOS) are computed using a geometrical sampling procedure. Once the DOS is obtained, the TSL is calculated using the LEAPR module of the NJOY code package. The calculated TSL of ZrH1.5 and ZrH2 are used in estimating the isothermal reactivity coefficient of the TRIGA Mark II research reactor at Jožef Stefan Institute and compared to experimental measurements.

07.09.2021 15:20 Reactor physics & research reactors I

Reactor physics & research reactors - 301

Application of an artificial neural network to support the design of the PWR reactor core configuration

Wojciech Kubiński, Patryk Bojarski, Piotr Darnowski

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

wojciech.kubinski@pw.edu.pl

The paper presents the use of the artificial neural network (ANN) as a tool supporting the design of the core configuration. The PWR reactor core, taken from the BEAVRS benchmark, was used to teach ANN to determine selected parameters of the core using the nodal diffusion core simulator PARCS. The network was used to determine the optimal configuration that attempts to meet the predetermined requirements (power peaking factor, effective multiplication factor, length of the cycle). Finally, the obtained configuration was discussed and compared with the initial model from the benchmark, determining the applicability of ANN in nuclear core design.

07.09.2021 15:40 Reactor physics & research reactors I

Reactor physics & research reactors - 303

Analysis of the X2 VVER-1000 benchmark with FENNECS

Romain Henry, Armin Seubert, Jeremy Bousquet

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

romain.henry@grs.de

In the last few years, a growing interest has been observed for small modular reactor (SMR). They are often characterized by complex geometry that classical nodal method cannot describe and, whereas Monte-Carlo (MC) methods are standard to describe the steady-state of such system, transient application at a reasonable ‘cost’ are not yet available. In order to perform multi-physics transient safety assessment for these new concepts, the Finite ElemeNt NeutroniCS (FENNECS) code is currently in development at GRS.
After a preliminary phase of verification FENNECS is mature enough to enter the validation phase. This phase should be separate in two groups of exercises. The first one aims to assess the performance of FENNECS on regular geometry problem and the second one on irregular geometry.
In this paper, FENNECS is used to perform the X2 Hot Zero Power (HZP) Benchmark (VVER-1000). Results of FENNECS are compared with a reference solution produced with the MC code SERPENT.

07.09.2021 16:20 Nuclear regulations

Nuclear regulations - 701

Technical and scientific support organizations as an independent layer of defense-in-depth in licensing of nuclear facilities

Mitja Uršič, Andrej Prošek, Leon Cizelj

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

mitja.ursic@ijs.si

Licensing of the nuclear facilities is described as a process with two independent layers of defense-in depth. The first layer is controlled by the licensee and its contractors, who are responsible for the safety case supporting the license (or license amendment) application. The second layer is controlled by the competent regulatory body, who is responsible to evaluate and to decide on the licensing merits of the safety case submitted by the licensee.

Traditionally, the competent regulatory bodies are seeking technical and scientific advice from the nuclear technical and support organizations (TSO). In such arrangement, the assessments of the safety case developed by the TSO are placed entirely in the second defense-in-depth layer, controlled by the regulatory body.

It is nevertheless possible to think of the safety assessments provided by the TSO as the third layer of defense-in depth. It is namely well known that additional layers of defense-in depth could significantly improve the safety of the underlying processes, if they are independent of the preexisting layers. If not independent, the additional layers may be detrimental, as they could introduce new failure modes resulting in simultaneous failures of multiple layers.

In the paper we explore some challenges that the independence of the TSO may bring to the regulatory process. In particular, some activities implemented at Jožef Stefan Institute, primarily active in research and higher education and also one among the several Slovenian TSOs, to strengthen the independence of the TSO are discussed and analyzed. These include some possible undesirable consequences of the competitive contracting of TSOs by the licensee, which is one of the features of the Slovenian nuclear regulatory framework. Strong long-term research and higher education program are shown to be essential to maintain and further develop the independence of the TSO through advancing the expertise, infrastructures, independence and credibility.

07.09.2021 17:00 Nuclear regulations

Nuclear regulations - 703

Ensuring Nuclear and Radiation Safety during Pandemic in Slovenia

Igor Grlicarev, Igor Sirc, Matjaž Podjavoršek, Helena Janžekovič, Tomi Živko, Andreja Peršič, Metka Tomažič, Darja Slokan Dušič, Neža Kompare, Janez Češarek

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

andreja.persic@gov.si

The corona virus pandemic has severely limited the ability of nuclear safety authorities to function as exercised for decades. Depending on the situation and the given conditions which were also a subject of changes in 2020-2021, the regulatory activities were reorganized and adjusted to the situation.
In the beginning of the first wave of the pandemic, the Slovenian Nuclear Safety Administration (SNSA) fully adapted to the situation by implementing all of its statutory obligations with approximately 20 % of the staff working in the office and the rest implemented teleworking. The interaction with the licensees and registrants was modified accordingly to the general rules of limiting the spread of virus. The inspection unit started to conduct very first virtual inspections. Also the physical inspections, i.e. on-site inspections, were implemented but they were modified to assure less social interactions. Later the SNSA temporarily postponed all on-site inspections, however exceptions were reactive inspections. The daily contacts with the Krško NPP were maintained via the communication means and in a similar way also with other authorized parties. The status of the NPP Krško operation was followed through daily reports and phone conversation with the plant staff as appropriate. In line with the legislation related to pandemic daily updates about the NPP status were prepared, covering any changes and restrictions related to pandemic measures as well as to the operational status.
The SNSA informed the ENSREG secretariat about the national preparation and contingency planning for the nuclear safety of nuclear installations under its regulatory control in scenarios of widespread disease epidemic or pandemic. The SNSA also shared information about the current measures in the country with the U.S. Nuclear Regulatory Commission and provided information exchange through the IAEA Incident Reporting System (IRS), as well as reported to the WENRA Reactor Harmonization Working Group. The SNSA also collaborated with the IAEA assuring exchange of lessons learned in pandemic.
After the first pandemic wave the SNSA placed the request to the Krško NPP to implement a comprehensive analysis of the impacts of the pandemic on the plant operation. The operator developed the internal Epidemic or Pandemic Response Plan, which introduces measures to manage potential risks of the rise of infections.
After another surge of COVID-19 cases throughout the country in early October 2020, the SNSA decided to introduce measures preventing the spread of the virus among its staff. At first 50% of the staff started teleworking, but after just ten days a modified scheme was adopted, where only the minimum number of the staff was present in the offices. Like during the first wave the rotating schedule for the office staff has been prepared for several weeks ahead. All the scheduled meetings at the SNSA headquarters have been postponed and all duty trips abroad have been cancelled. The regular services regarding licencing and registration are being carried out online or via mail and this became the normal way of doing business.
The lessons learned of the SNSA activities during the proclaimed epidemic of the coronavirus will be presented in the article, focusing on the effectiveness of supervision and inspections of nuclear facilities in the country. The experience of the SNSA's preparedness for hypothetical nuclear events during an pandemic will also be given.

08.09.2021 09:10 Severe accidents II

Severe accidents - 405

Assessment of different combustion models for the numerical simulations of the THAI containment

Javier Lobato-Perez, Daniele Dovizio, Ed Komen

NRG , Westerduinweg 3, Postbus 25, 1755 ZG Petten, Netherlands

dovizio@nrg.eu

In the context of severe accident scenarios in a nuclear power plant, hydrogen management is an important component to ensure the reliability and proper functioning of critical systems in the case of a severe accident. In particular, if the concentration of hydrogen reaches flammability limits, combustion or explosion of hydrogen-air mixtures in the containment can damage crucial safety systems and can even compromise the safety function of the containment walls.
CFD codes can serve as a numerical tool for the assessment of the risks associated to hydrogen combustion during severe accidents, by providing detailed transient predictions when compared to one-dimensional system codes.
In the past, we validated CFD based methods to determine the pressure loads from a fast deflagration as well as a slow deflagration with and without the presence of steam. However, the results demonstrated that further research on laminar flame speed as well as its effect on the flame front propagation and flame development was needed, especially for lean hydrogen mixtures.
The objective of the present work consists of assessing the effects of a new correlation for the flame speed in order to further develop and validate the in-house code of NRG when dealing with hydrogen combustion for safety management analysis. Furthermore, it has been observed that the new correlation predicts large values of the flame speed near the wall. In order to make the model as general as possible and to implement no user defined parameters to cut off the wall distance, a wall bounded version of the new combustion model is introduced, by means of a blending function similar to the one used in the ? – ? SST turbulence model.
Different combustion models are validated for slow deflagrations in a closed containment, namely the Thermal hydraulics Hydrogen Aerosol and Iodine (THAI) facility. Effects of flame propagation and buoyancy forces are investigated.
Numerical results are presented and analyzed in terms of flame front development and propation and pressure rise. Overall, it can be stated that the new combustion model results in a more robust approach: it presents no special requirements on the maximum Courant number or turbulence model, while achieving flame propagation in all cases and providing good predictions in terms of flame evolution and pressure rise. Furthermore, the other combustion model show high sensitivity of the results on the initial conditions.

08.09.2021 09:30 Severe accidents II

Severe accidents - 406

Numerical Simulation of Melt Droplet/Water Interaction in Steam Explosions

Nikita Sivakov1, Sergey E. Yakush2

1Bauman Moscow State Technical University, ul. Baumanskaya 2-ya, 5/1, Moscow, 105005, Moscow, Russian Federation

2Institute for Problems in Mechanics Russian Academy of Sciences (IPM RAS), pr. Vernadskogo 101, Moscow, Russian Federation

yakush@ipmnet.ru

Numerical simulations of melt droplet-water interactions relevant to the steam explosion phenomenon are performed by the Volume-of-Fluid (VOF) method. Initially, a spherical high-temperature melt droplet is considered with a thin vapor film separating it from the surrounding subcooled water. The steam explosion wave is modeled by sudden increase of pressure in water, causing development of instability in the vapor film and its collapse, with direct impingement of water on the melt surface. The mathematical model considers the melt and water as weakly compressible viscous liquids, whereas the water vapor properties are described by the IAPWS steam tables. Three-dimensional governing equations are solved on a multiblock Cartesian mesh refined adaptively to resolve the interfaces between the phases. It is shown that water impact on the melt surface causes significant perturbation of liquid droplet, resulting in the formation of fast small-scale melt jets and droplets penetrating in water. Simulations are performed in wide ranges of the initial pressure in water, droplet diameter, vapor film thickness, melt temperature and water subcooling. The integral parameters of the thermal interaction obtained numerically include the time histories for the melt surface area, maximum pressure in the domain, and total evaporation rate. Simulation results are compared with the experimental data on single droplet steam explosions available in the literature. Implications for the problem of melt fragmentation in the steam explosion wave are discussed.

08.09.2021 09:50 Severe accidents II

Severe accidents - 413

EMPLOYMENT OF THE ASTEC CODE IN THE SEVERE ACCIDENTS RESEARCH ACTIVITIES AT KIT

Fabrizio Gabrielli1, Victor Hugo Sanchez1, Walter Tromm2

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

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

walter.tromm@kit.edu

The improvement of the performance of the modern integral codes for carrying out severe accident analyses in NPPs is one of the milestones of the Nuclear Safety Research program (NUSAFE) of the Karlsruhe Institute of Technology (KIT). Such activity is performed in the frame of the participation of KIT to European projects and international cooperation on the safety assessment of NPPs including innovative designs, e.g. SMRs, and also aims at supporting decision making under high uncertainties for all emergency situations.
The KIT strategy on the employment of integral codes is based on a clustering the computational issues (development and modelling) with the free access to the in-situ severe accident large infrastructures, e.g. QUENCH, LIVE, HYKA. Such unique research environment, allows performing severe accident research in multiple directions: V&V, development of advanced physical models and mathematical methods, uncertainty quantification, and analysis.
In this framework, the use of the Accident Source Term Evaluation Code (ASTEC), developed by IRSN, plays a central role to analyse the complete SA scenario from the initiating event until radioactive release from the containment in Gen. II and Gen. III water-cooled reactors.
In this paper, the results of recent KIT activities devoted to the ASTEC code validation and to its application for severe accident plant analyses are shown. The QUENCH-08 (PWR bundle) and QUENCH-20 (BWR bundle) are considered for ASTEC validation. The main goal of such tests is analysing the physical phenomena occurring during water injection into a partially degraded core, which is one of the prime SAM measures to prevent the failure of the safety barriers i.e. Reactor Pressure Vessel. In fact, under particular conditions, steam generated during reflooding may significantly enhance the Zircaloy cladding oxidation accompanied by temperature escalation and then trigger a spiky hydrogen generation. The analysis of the results shows that ASTEC is able to reproduce the experimental data of hydrogen production and cladding oxide thickness in instrumented bundle positions. Namely, the code is able to predict the key phenomena governing the effects of reflooding in the early in-vessel phase of a severe accident in PWRs and BWRs.
In line with the KIT strategy for nuclear reactor safety, the plant applications are focused on the evaluation of the Source Term during a severe accident scenario. The ASTEC results of a Medium Break (12’’) Loss of Coolant Accident with and without Station Black Out for a generic KONVOI-1300 NPP are discussed in the paper. The results show the capability of ASTEC to simulate the full scenario, namely simulating the key in-vessel and ex-vessel phenomena as well as their effects on the transport of the Fission Products released by the core to the plant and the environment. In particular, the peculiarity of the ASTEC code to be able to evaluate the element- and isotope-wise composition of the radiological release is of relevance in view of the assessment of a Source Term database to be employed by predictive tools, e.g. JRODOS.
Such validation and analysis activities for ASTEC are therefore of relevance since they integrate the main phenomena affecting a severe accident and the following radiological impact.

08.09.2021 11:00 Reactor physics & research reactors II

Reactor physics & research reactors - 304

Comparing Different Approaches to Calculating Decay Heat Power of a Spent Fuel Dry Storage Cask for Krško NPP

Vid Merljak1, Marjan Kromar2

1Krško Nuclear Power Plant, Vrbina 12, 8270 Krško, Slovenia

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

vid.merljak@nek.si

One of the main limitations for dry storage of spent nuclear fuel is its decay heat power. Direct measurements are quite rare since they are time-consuming and expensive to perform. Therefore, computational approaches have been devised in the past to calculate the decay heat power. We can distinguish at least three approaches: 1) using (semi-)empiric formulae; 2) physics calculation while grouping fuel assemblies with similar characteristic and using only the most limiting value of each parameter (the so-called bounding approach); or 3) best-estimate calculation using explicit data of each fuel assembly.
In this paper, we compare results of such calculations for the case of a single dry storage cask with 37 fuel assemblies from the Krško NPP. Best-estimate calculations were run with the ORIGAMI Automator (OA) of SCALE 6.2.2 code system, while the fuel assembly data was taken from an official Fuel Assembly Register (FAR) database. Due to data-intensive and error-prone input to OA project, a Python script interface FAR2OA was made and is briefly described here. Final results of decay heat power comparison show that the calculation approaches agree to a reasonable extent. Thus, FAR-FAR2OA-OA sequence is verified as successful.

08.09.2021 11:20 Reactor physics & research reactors II

Reactor physics & research reactors - 305

On the Processing of Thermal Scattering Law Data into ACE Format for Monte Carlo Transport Calculations

Daniel Lopez Aldama1, Andrej Trkov2, Roberto Capote1

1International Atomic Energy Agency, Wagramerstr. 5, P.O.Box 100, A-1400 Vienna, Austria

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

andrej.trkov10@gmail.com

In recent years several codes for producing application libraries from evaluated nuclear data files became available, in addition to NJOY, which is the standard code for producing libraries in ACE format for Monte Carlo transport calculations at the Los Alamos National Laboratory. A code verification exercise was conducted through the International Atomic Energy Agency (IAEA) to test the capability of the codes to generate ACE libraries in the fast energy range. The second phase was started recently to test the capability of the codes to process thermal scattering law data (TSL) into ACE_TSL format. The TSL of Zrh were chosen as an example. The data were tested on a number of criticality benchmarks from the International Criticality Safety Benchmark Experiment Project (ICSBEP). Different format options exist for representing the TSL data in the ACE_TSL files. In this work the focus is on the ACEMAKER code, with which a number of sensitivity studies were made. The results were found to be sensitive to the incident energy mesh by as much as 0.1% in k_eff due to data processing alone. Some results of the sensitivity to the data processing options will be presented and recommendations for implementation in the processing codes will be made.

08.09.2021 11:40 Reactor physics & research reactors II

Reactor physics & research reactors - 306

Upgrades to the Monte Carlo Computational Model of the JSI TRIGA Research Reactor

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

anze.pungercic@ijs.si

TRIGA Research Reactor at the “Jožef Stefan” Institute is one of the most utilized research reactors due to the ability to perform versatile experiments, which is a direct result of having well characterized neutron and gamma field in every part of the reactor core and its surroundings. This was achieved in last 20 years by the detailed development of the TRIGA’s computational model in the MCNP Monte Carlo neutron transport code and its extensive validation on different experiments and benchmarks. This model includes detailed geometry of the reactor core with its surrounding with fresh fuel isotopic composition at room temperature. Such description was sufficient for neutron spectra, reaction rate profiles and gamma field calculations However the criticality was consistently overestimated, resulting in discrepancies between calculated and measured reactivity of approx. 5000 pcm.. We investigated the reasons for this discrepancy and decided to improve the model by including burnup and temperature effects to the model.

Main objective of this paper is to thoroughly analyze the origin of thediscrepancy between measured and calculated core reactivity and upgrade the existing computational model. One of the highest contributors to the discrepancy was found to be fuel burnup. We used the isotopic composition, obtained by detailed burnup calculations using the Serpent 2 code, to determine the effect of fuel burnup to be around 4500 ± 500 pcm (uncertainty is due to operational history modelling and axial burnup distribution due to control rod insertion). The remaining ~ 1000 pcm discrepancy is connected to fuel temperature, xenon build-up and inelastic S(a,b) cross section of hydrogen bound in zirconium. Each individual source of keff discrepancy will be further discussed and presented in the final paper. In the end the upgraded computational model was used to repeat the calculation of 197Au(n,?)198Au and 27Al(n,?)24Na reaction rate axial distributions and compare it to the experimental values. The agreement between calculations and experimental values improved. Hence improved model will be used in the future to support experimental campaigns.

08.09.2021 12:00 Thermal-hydraulics & CFD II

Thermal-hydraulics & CFD - 604

Experimental exploration of interweaved fluctuations of pressure and velocity in 5x5 rod bundle with spacer grids

Naz Turankok1, Lionel Rossi2, Valerie Biscay2, Thibaud Lohez2

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

naz.turankok@cea.fr

In Pressurized Water Reactor (PWR) core, fuel assemblies are made of fuel rods maintained by spacer grids. These spacer grids have different elements such as dimples, springs (to support the fuel rods) and possibly mixing vanes to enhance mixing. With the aging of the spacer grids, fluid-structure interactions provoke vibrations leading to fretting by contacts between the rods and grid elements. This fretting is one of the major reason for the degradation of the outer layer of the fuel rods. Within the turbulent rod bundle flow, the spacer grids are an additional source of turbulence. The experimental low-pressure water rig, named CALIFS, consists in a 5x5 rod bundle maintained with analytical spacer-grids. It has optical accesses with Perspex windows on three sides. The working ranges are 0-400m3/h for flowrate and 10°C-55°C for the temperature. Using CALIFS, (Turankok et al. NED 2019) show the existence of frequency peaks in pressure fluctuations spectra downstream spacer grids with and without mixing vanes for a Reynolds number of 66 000. In a recent article, using rods with FEP parts to perform PIV measurements close to the central instrumented rod, (Turankok et al. EIF 2021) highlight the flow phenomena plausibly responsible for the frequency peaks. New Reynolds numbers from 13000 to 108000 are explored using spacer grids without mixing vanes. The presence of coherent vortex streets is now supported by LDV and PIV measurements. Current work includes the quantification of the transport of pressure fluctuations along with simultaneous pressure and PIV measurement using multi-pressure sensors devices. These results shown that the pressure fluctuations are convected with the speed of flow. The correlations between pressure fluctuations and the velocity fluctuations (and derivatives, e.g. vorticity, swirling functions) are currently explored using simultaneous measurements data.

08.09.2021 12:20 Thermal-hydraulics & CFD II

Thermal-hydraulics & CFD - 605

Scale effects methodology for buoyant impinging jets in Sodium Fast Reactors

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

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

4LEMTA Université de Lorraine CNRS , 2, av. de la Foret de Haye - TSA 60604, 54518 Vandoeuvre les Nancy, France

benjamin.jourdy@cea.fr

The CEA is involved in the development of 4th generation sodium-cooled fast 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. A critical issue in SFR reactor is the rising of the jets outgoing the core at low operating power: this phenomenon induces thermal fluctuations in the vessel, leading to thermal fatigue of the components. Therefore, the scale effect study is focused on the behaviour of these jets, especially on the buoyancy effect becoming preponderant when inertia is decreasing.
Previous studies on this issue were already performed on a 1:6 scaled mock-up representative of a SFR’s hot plenum: MICAS. Theoretical scale studies based on the Vaschy-Buckingham theorem and dimensionless Navier-Stokes equation under Boussinesq’s approximation led to relevant dimensionless numbers such as the densimetric Froude number and the Euler number.
The scale effect methodology based on the scale mock-up series’ is detailed, using MICAS as a reference scale. A new mock-up called MOJIT-Eau has been designed to be representative of the smaller scales: this mock-up can be modified to investigate scale effects from 1:4 to 1:2,5 of MICAS’s scale, and allow main phenomena investigation. The objective is to quantify the influence of the complex geometry on the jets raise. Due to the size reduction, some geometrical distortion has been made in order to ensure the turbulence of the flow and also to quantify the relative influence of the Euler number compared to the densimetric Froude number.
This mock-up is in a transparent polymer vessel to carry out optical measurements as PIV. Temperature fields are measured thanks to optical fibre and PT100 probes in the vessel. Measurements will lead to experimental dimensionless numbers in order to compare their relative influence on the jets’ raise.

08.09.2021 12:40 Thermal-hydraulics & CFD II

Thermal-hydraulics & CFD - 606

Multiphysics Topology Optimization with Application to Molten Salt Fast Reactor

Lorenzo Cattoni1, Carolina Introini2, Antonio Cammi3, Laura Savoldi4, Rosa Di Fonzo4

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

2Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

3Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy

4Politecnico di Torino, c.s. Duca degli Abruzzi 24, 10129 Torino, Italy

carolina.introini@mail.polimi.it

In the last decades topology optimization is playing an increasing important role in the industrial design approach for different applications including structural mechanics, civil engineering, architecture and fluid mechanics. This work was aimed at improving the studies regarding multiphysics topology optimization of systems governed by fluid flow and heat transfer including an application to the Molten Salt Fast Reactor (MSFR). Two different optimization approaches were investigated. First, the well-established topology optimization gradient based algorithms were studied with 2D benchmarks using the optimization module available in COMSOL Multiphysics and in STAR-CCM+. These results represented the starting point for the development of open source optimization solvers in OpenFOAM for flows including heat transfer based on the adjoint approach. This method, defined by a Lagrangian formalism of the sensitivity analysis, led to promising results characterized by lower computational cost but higher residuals which underlined the need of further validation tests. Then, the EVOL geometry of the MSFR was subjected to topology optimization aimed at minimizing the temperature gradient and pressure drops inside the reactor. The results showed impressive improvements regarding the system operative conditions.

09.09.2021 09:10 Nuclear fusion and plasma technology

Nuclear fusion and plasma technology - 801

On the potential of W2C-reinforced tungsten: microstructure and mechanical properties of samples aged at temperatures above 1000 °C

Petra Jenuš1, Matej Kocen1, Anže Abram1, Aljaž Ivekovič2, Črtomir Donik3, Irena Paulin4, Saša Novak5

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

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

3Inštitut za kovinske materiale in tehnologije, Lepi pot 11, 1001 Ljubljana, Slovenia

4The Institute of Metals and Technology, Lepi pot 11, 1000 Ljubljana, Slovenia

5Jožef Stefan Institute, Department for nanostructured materials, Jožef Stefan International Postgraduate School, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

petra.jenus@ijs.si

Tungsten has good high-temperature physical properties, which has made it a material of choice for the ITER divertor. On the other hand, tungsten is also associated with a reduction in its mechanical properties related to its grain growth at and above its recrystallization temperature (above 1000 °C) [1]. The quest to improve the tungsten’s properties to sustain plasma-facing conditions in the DEMO divertor is still undergoing. Among the available options, we selected the reinforcement of tungsten with ditungsten carbide nanoparticles (W2C) since the reinforcement does not chemically react with the matrix [2]. The aim of this study is to present whether ageing at elevated temperatures (above 1250 °C) affects the microstructure and mechanical properties of W-W2C composites.
Carbide particles in W-xW2C composites (0 ? x ? 40 wt%) were formed in-situ during the thermal treatment of powder mixtures consisting of W and WC particles at various ratios with a field-assisted sintering technique (FAST) at 1900 °C with a heating rate of 100 °/min, applied pressure of 60 MPa and holding time at final temperature 10 min. As-sintered samples were aged at 1250 °C, 1600 °C and 2000 °C for 24 h in a vacuum. Both as-sintered and aged samples were inspected in terms of phase composition, microstructure, grain size and mechanical properties (Vickers hardness).
The microstructural analysis confirmed that in the as-sintered samples, the W2C precursor enhances densification by the reduction of surface oxide and that the higher carbides' concentration leads to smaller grains of the as-synthesized composites. The study showed that during ageing W2C grains stabilize the microstructure of W-W2C composites.

Acknowledgement: 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] Rieth, M.; Dudarev, S. L.; Gonzalez de Vicente, S. M.; Aktaa, J.; Ahlgren, T.; Antusch, S.; Armstrong, D. E. J.; Balden, M.; Baluc, N.; Barthe, M.-F.; et al. Recent Progress in Research on Tungsten Materials for Nuclear Fusion Applications in Europe. J. Nucl. Mater. 2013, 432, 482–500.
[2] Novak, S.; Kocen, M.; Šestan Zavašnik, A.; Galatanu, A.; Galatanu, M.; Tarancón, S.; Tejado, E.; Pastor, J. Y.; Jenuš, P. Beneficial Effects of a WC Addition in FAST-Densified Tungsten. Mater. Sci. Eng. A 2020, 772.

09.09.2021 09:30 Nuclear fusion and plasma technology

Nuclear fusion and plasma technology - 802

Liquid metal divertor in future fusion reactors: liquid tin

Rok Zaplotnik1, Vincenc Nemenič2, Janez Kovač2, Miran Mozetič3

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

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

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

rok.zaplotnik@ijs.si

Production of energy without greenhouse gas emissions in the future is a necessity. One part of the solution for this problem are fusion power plants. The advantages of fusion are: sustainability, no CO2, no long-lived radioactive waste, limited risk of proliferation, no risk of meltdown, and there is abundant fuel.
There are many fusion reactors around the world, currently the largest being JET (Joint European Torus), where fusion experiments are performed. However, none of these reactors can produce more power than it consumes, the first reactor that will be able to do that is being built in the south of France. The ITER tokamak will be a unique experimental tool capable of producing ten times the return of invested energy. However, ITER will not yet be a power plant, but it will contribute to the design of the next-generation machine, a demonstration power plant DEMO.
The divertor, which is positioned at the bottom of the vacuum vessel, controls the exhaust of waste gas and impurities from the reactor and withstands the highest surface heat loads of the tokamak. In ITER, the divertor is made from tungsten, but in DEMO this will not necessarily be the case. The final design of the divertor in DEMO is not yet determined.
For DEMO and beyond, also a liquid metal plasma-facing components are being considered [1]. Only a few metals and alloys with a low melting point have been recognized as suitable candidates, like for example tin, lithium, gallium, LiSn, etc. Currently, tin seems to be the most promising candidate. In order to research the tin properties, experiments are being performed in large scale experiments, such as fusion tokamak COMPASS, and Magnum-PSI [1]. Complementary to those large scale experiments also smaller laboratory experiments are performed in plasma labs such as ours in the Department of surface engineering at Jožef Stefan Institute. Our study focused on hydrogen solubility and neutral D atom retention in liquid tin will also be presented.

[1] Morgan T. W., Rindt P., G van Eden G., Kvon V., Jaworksi M. A.and Lopes Cardozo N.J., Plasma Phys. Control. Fusion, 60 (2018) 014025.

09.09.2021 09:50 Nuclear fusion and plasma technology

Nuclear fusion and plasma technology - 803

Assessment of Power Deposition on Plasma Facing Components Inside WEST Tokamak With the Use of Field Line Tracing

Matic Brank1, Gregor Simič2, Leon Kos1, Mehdi Firdaouss3, Marie-Helene Aumeunier3

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

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

3Institute for Magnetic Fusion Research, CEA, Cadarache, 13115 Saint-Paul-lez-Durance, France

Assessment of plasma power deposition with the use of magnetic field line tracing on tokamak plasma facing components is important in order to provide thermal loading of critical components in the fusion reactors. Such analysis can provide insight into PFC design and can improve the durability and safety of nuclear fusion reactors.

This article describes a benchmark performed between two magnetic field line tracing codes, PFCFLUX[1] and SMITER[2]. The code PFCFLUX (Plasma Facing Components Flux) has been developed for heat flux calculations on those components, including shadowing effects. SMITER is a graphical user interface (GUI) framework, built around SMARDDA kernel for magnetic field line tracing and power deposition mapping on tokamak plasma-facing components (PFC) in the full 3-D CAD geometry of the machine. The main purpose of the benchmark presented here is the evaluation of the power deposition on the poloidal limiters (high and low field sides) and the divertor targets of WEST tokamak. Magnetic coils of WEST have been added to confine the originally circular plasma into and ITER-like "D"-shape. As one of the main testing tokamaks for ITER, it is thus also suitable to provide assessment of heat loads on plasma facing components inside WEST.

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 is considered non-wetted. If the field line does not hit anything after a user-defined maximum distance, then the target triangle 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.

[1] M. Firdaouss et al., Modelling power deposition on the JET ITER like wall using the code PFCFLUX, Journal of Nuclear Materials, 438, 2013, 536--539

[2] L. Kos et al., SMITER: A field-line tracing environment for ITER, Fusion Engineering and Design, 03, 2019, DOI: 10.1016/j.fusengdes.2019.03.037

09.09.2021 11:00 Materials and ageing management

Materials and ageing management - 901

A simplified model for estimating intergranular normal stresses

Samir El Shawish, Timon Mede

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

timon.mede@ijs.si

Due to the widespread use of metallic alloys in various industries the ability to predict the most significant ageing and material degradation modes in them is vital. Particularly critical in this respect is the process of intergranular stress-corrosion cracking (IGSCC) which causes the development of cracks along the grain boundaries (GBs) in polycrystalline aggregates under external macroscopic loading ?. Its initiation depends predominantly on the intergranular normal stresses (INS) and the strength of the corresponding GBs. Those can be classified into different types associated with different IGSCC sensitivities. A statistical correlation exists between a specific GB type and the INS distribution ?nn on it. Wider distributions in general imply a larger fraction of highly stressed GBs at the same applied load and thus increased probability for local GB stress exceeding a critical value (related to GB strength). INS distributions can be characterized by their first two statistical moments, the mean value <?nn/?> and standard deviation s(?nn/?).

Numerical finite element simulations demonstrate that the statistical behaviour of GB normal stresses in materials with elastic cubic grains can be accurately described for any GB type and loading condition by just two parameters: Zener index A which characterizes the material elastic anisotropy and the effective stiffness E12 of the two grains surrounding the GB. The newly introduced E12 parameter specifies a GB type by combining its geometrical aspect with its material properties. It measures the average stiffness of GB's immediate neighbourhood along the GB normal direction. The amplitude of INS fluctuations measured by s(?nn/?) reduces if both grains are softer than the surrounding material and thus the applied stress projected along the GB normal redistributes more over the stiffer bulk than over the softer grains. Conversely the largest normal stresses most likely form on GBs whose normals are oriented along the stiffest direction in both adjacent grains.

To provide an intuitive explanation for our findings a simplified bicrystal model embedded in an elastic medium was introduced in which the significance of effective Young’s modulus E12 can be explicitly shown. Even though the phenomenological relation between the INS fluctuation amplitude and the corresponding GB type has been derived assuming uniaxial tensile loading, similar relations can be formulated even for other types of loading (such as general triaxial loading presented here) with the characteristic parameter E12 still playing a pivotal role. Analytical form of INS allows us to quickly estimate INS distributions without plunging into numerical simulations. The actual shape of ?nn/? distributions (and not just the corresponding standard deviations s(?nn/?) of normalized INS) are needed to measure the fraction of GBs with large normal stresses. Comparison with finite element results show that the proposed bicrystal model is able to accurately predict the INS distributions for random GBs as a function of both elastic anisotropy A and loading conditions but only recovers the correct trends for their standard deviations when restricted to individual GB types characterized by E12. These results are especially relevant for understanding IGSCC initiation and potentially engineering materials with improved cracking resistance.

09.09.2021 11:20 Materials and ageing management

Materials and ageing management - 902

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Dmitry Terentyev1, Saša Novak2, Petra Jenuš3, Carmen Garcia-Rosales4, Elisa Sal4, Jan Willem Coenen5

1SCK CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 MOL, Belgium

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

4Ceit Centro Tecnológico, Paseo de Manuel Lardizábal, No 15, 20018, Donostia - San Sebastián, Spain

5Forschungszentrum Jülich, Wilhelm-Johnen-Straße, 52428 Jülich, Germany

dterenty@sckcen.be

Development of refractory metals for application as plasma-facing armour material remains among priorities of fusion research programmes in Europe, China and Japan. Improving the resistance to high temperature recrystallization, enhancing material strength to sustain thermal fatigue cracking and tolerance to neutron irradiation are the key indicators used for the down selection of materials and manufacturing processes to be applied to deliver engineering materials.
In this work we investigate the effect of neutron irradiation on mechanical properties and microstructure of several tungsten grades recently developed in Europe and China. Neutron irradiation campaign is arranged for screening purposes and therefore is limited to the fluence relevant for the ITER plasma facing components. At the same time, the neutron exposure covers a large span of irradiation temperatures from 600 up to 1000 °C. Four different grades are included in the study, namely: fine-grain tungsten strengthened by W carbides (W-4wt.% W2C), fine-grain tungsten strengthened by Zr carbides (W-0.5% ZrC), W alloyed with 10 at.% chromium and 0.5 at.% yttrium (W-10Cr-0.5Y) and technologically pure W plate manufactured according to the ITER specification by AT&M (China). The strengthening by W2C and ZrC particles leads to an enhanced strength, moreover, the W-0.5ZrC material exhibits reduced DBTT (compared to ITER specification grade) and is available in the form of thick plate (i.e. high up-scaling potential). The W-10Cr-0.5Y grade is included as the material offering the self-passivation protection against the high temperature oxidation. The results are presented and discussed across the currently available knowledge on the irradiation effects in pure tungsten and other refractory metals.

09.09.2021 11:40 Materials and ageing management

Materials and ageing management - 903

Ageing at the TRIGA MK II research reactor of the University of Pavia: management and practical applications

Andrea Gandini1, Anna Maria Condino2, Andrea Salvini1

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

2University of Pavia, Dept of Drug Sciences, Via Taramelli 12, 27100 Pavia, Italy

andrea.gandini@unipv.it

In the Laboratory of Applied Nuclear Energy (LENA) of the University of Pavia a 250 kW TRIGA Mark II research reactor is in operation since 1965 without unplanned extended shutdown periods during which different activities took place. Ageing and its management, in a more than fifty years old facility, is a key point for both reactor safety and all its activities continuation. For these reasons, to mitigate ageing effects, several issues must be considered due to the time-dependent degradation of TRIGA and LENA structures, systems, and components (SSCs). Furthermore, from a management point of view, the centre must deal with typical problems as documentation control, changes in mandatory requirements and problems in staff turnover, that could lead to a loss of experience and know-how. During the past years, starting from an accurate assessment of SSCs conditions and the identification of ageing mechanisms, several activities were successfully carried out. In this field, LENA has implemented, since 2014, an ageing management system, continuously revised until its final transmission to the Italian control body. From the experience gained over these years, a new improvement phase has begun to identify the difficulties in the plan application and the opportunity to integrate changes to make it more practical and effective.
In this paper, an overview of the above-mentioned topics, and the forthcoming plans highlighting lessons learned and challenges, will be provided.

09.09.2021 12:00 Environment and back end of the fuel cycle I

Environment and back end of the fuel cycle - 1019

NPP Krško Low and Intermediate Level Waste: Is More Rational Economical and Financial Management Plan Still Possible?

Alemka Knapp, Ivica Levanat, Diana Šaponja-Milutinović

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

ilevanat@tvz.hr

According to the recently prepared 3rd Revision of the Krško NPP Radioactive Waste and Spent Fuel Disposal Program, low and intermediate level waste (LILW) will be divided in equal parts to be separately managed by Slovenia and Croatia.
After several years of negotiation, disposal at the Slovenian LILW repository in Vrbina was not agreed upon as a joint solution. The Intergovernmental Agreement requires that each country removes its share of waste from the NPP premises by 2025 – unless the joint solution is found by 2023, which now appears most unlikely.
However, full implementation of separate LILW management by 2025 does not appear much more likely either. To say nothing about rational aspects of the national LILW management plans.
Construction of the Slovenian repository (not yet really started) will be rather demanding and expensive (and certainly cannot be quite completed before 2025). And even more expensive are the compensations to the local community, already being paid for about a decade.
Development of the Croatian radioactive waste management center is at the stage of initial natural background radiation measurements, on the prospective site in a hostile community. Establishment of a simple surface storage facility, in the first phase, might be completed within several years (though not as early as 2013), at worst by circumventing the local community consent by the State issued administrative provisions.
And yet, it would not be entirely impossible to remove the LILW from the NPP “somewhere around 2025”.
Slovenia can move its share to the nearby Vrbina disposal site even before serious repository construction works begin, and carry out any remaining treatment activities on that site.
Croatia plans to send its share to a third country, for conditioning and packaging. The shipments can begin and proceed even before the Croatian storage facility is completed.
However, given the fact that many details of the waste division and take-over (legal, technical and practical) have not yet been fully clarified, it is highly unlikely that the LILW removal will be fully completed in 2025. Therefore, both countries will probably tacitly tolerate minor departures from the Intergovernmental Agreement, rather than attempt any small modifications.
The authors, on the contrary, propose that Slovenia and Croatia undertake an explicit modification of the Intergovernmental Agreement before 2023, in order to postpone the 2025 waste removal deadline for at least a decade or more. Such modification
- would eliminate the need for hasty implementation of the present management plans, in which optimal solutions might be overlooked,
- may facilitate financing of these plans, through postponement of major expenses, and
- could even open possibilities for serious revisions of present plans, so that at least some elements of joint management might be included.
The authors are convinced that present plans are not very rational solution for management of small LILW quantities from a single medium NPP.
The paper discusses the outlined potential advantages of the proposed Agreement modification, arguing that more rational LILW management options may still be possible.

09.09.2021 12:20 Environment and back end of the fuel cycle I

Environment and back end of the fuel cycle - 1001

The French Reprocessing Solution And Its Recent Evolution

Albert Yokobayashi

ORANO, 1 Place Jean Millier, 92400 Courbevoie, France

albert.yokobayashi@areva.com

Used fuel reprocessing offers several benefits for operators, reducing risks and costs, enhancing public support and allowing safe and long-term storage of final waste as well as geological repository rationalization.

The Challenges to Used Fuel Management

Many utilities and waste management organizations in charge of used fuel management face several challenges such as saturation of the pools, delays of final disposal projects, needs of multiple operations related to the management of long term interim storage and downstream steps (Potential issues with fuel integrity for extended dry storage period, ageing management of storage systems with potential required reconditioning, conditioning of used fuel into final disposal canisters…). In this context, taking into consideration requirements from local or international authorities to implement a used fuel management program up to waste disposal, used fuel reprocessing provides a solution to implement a safe, cost effective and sustainable strategy. It allows an optimized conditioning of final waste and the reuse of recyclable materials contained in used fuel, thus saving natural uranium resources and increasing the energetic independence.

The Reprocessing Solution

Used fuel reprocessing is a mature technology industrially implemented for decades by international nuclear operators leading to an optimized final waste conditioning.
Reprocessing performed in the Orano La Hague facility for over 40 years includes:
• extraction of the recyclable materials (U & Pu) from the used fuel, representing 96% of nuclear materials contained in used fuel;
• conditioning of the non-recyclable materials considered as final waste (fission products, minor actinides and non-soluble metallic parts) in Universal Canisters, respectively vitrified and compacted (Universal Canister of Vitrified Waste, UCV and Universal canister of Compacted Waste, UCC).
Advantages of such Universal Canisters :
• long-term stability of the High-level waste (vitrification of fission products and Minor actinides) conditioned in a glass matrix specifically designed for long term storage, transport and durability under long term geological conditions (over thousands of years);
• standardization of the final waste form with metallic compacted waste and vitrified waste conditioned in same Universal Canister and compatibility with geological disposal projects such as the Andra’s underground radioactive waste repository project, Cigeo, in Bure (France) ;
• removal of fissile material from HLW leads to a reduction of radiotoxicity of the final waste and alleviate the IAEA safeguards requirements
• optimization of disposal facility design with simplification of handling operations, reduction of the volume needed for disposal
• a globally strong risk reduction approach for the management of final waste.
Recent evolution of the French Reprocessing solution
In France, a set of decrees have been recently issued by the government allowing the return of the global inventory of UCC and UCV from two countries of origin using different ratio of UCV and UCC for each country, provided that all metallic mass and radioactive activity of both countries leaves France. This flexibility is based on the calculation of the "Integrated Toxic Potential” representative of the waste canisters radiotoxicity which needs to be agreed between the two countries and Orano.

This presentation will focus on this new approach integrating a flexibility in the system of attribution for UCC and UCV providing additional value to reprocessing clients.

09.09.2021 12:40 Environment and back end of the fuel cycle I

Environment and back end of the fuel cycle - 1002

Paulina Dučkić, Davor Grgić, Siniša Šadek, Štefica Vlahović

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

paulina.duckic@fer.hr

Nuclear power plant Krsko opted for Spent Fuel Dry Storage (SFDS) to ensure place in the Spent Fuel Pool (SFP) for the additional 20 years of prolonged lifetime. Spent Fuel Assemblies (SFAs) will be relocated from the SFP to the SFDS in four loading campaigns. In total, 2294 SFAs will be placed in a dry storage building arranged in 62 HI-STORM FW casks, 37 SFAs in each cask.

09.09.2021 14:20 Education, training and outreach

Education, training and outreach - 1101

Strategic agenda for EU wide nuclear education, training and knowledge management

Leon Cizelj1, Csilla Pesznyak2, Michele Coeck3, Joerg Starflinger4, Pavel Gabriel Lazaro5, Franck Wastin6

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

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

3SCK CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 MOL, Belgium

4Institute of Nuclear Technology and Energy Systems (IKE) University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart, Germany

5European Nuclear Education Network, Rue d’Egmont 11, 1000 Brussels, Belgium

6European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands

leon.cizelj@ijs.si

The technical, regulatory and market complexities that govern the development, operation, decommissioning, waste management and oversight of nuclear power plants require personnel with outstanding knowledge, skills and motivation. Apart from being technical specialists understanding the installations of increasing technical complexity, they must be prepared to work in increasingly multidisciplinary, multicultural and highly competitive environments.
One would therefore expect that the attraction of new talents followed by the high-class education of nuclear professionals is a thriving activity enjoying strong support by all stakeholders.
An EU wide strategic agenda for nuclear education, training and knowledge management is being developed within the ENEN+ project to support and consolidate the efforts of the nuclear stakeholders to attract, develop and retain new talents. On the one hand, it is based on the projections of needs developed by the European Human Resources Observatory-Nuclear (EHRO-N). On the other hand, it builds on the existing national nuclear education strategies in EU and beyond and nearly two decades of experience of European Nuclear Education Network (ENEN aisbl).
The paper will outline and substantiate the most important actions required for the EU to retain its leading role in the nuclear reactor engineering and safety, geological disposal, radiation protection and medical applications. The sheer complexity of this challenge calls for high level of support, coordination and partnership between all nuclear stakeholders, especially those involved in all levels of decision-making.

09.09.2021 14:40 Education, training and outreach

Education, training and outreach - 1102

Socio-economic impacts of nuclear power plant closure: Lessons learnt from different EU Member States

Hana Gerbelova, Vaida Rukaite Drazdove

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

hana.gerbelova@ec.europa.eu

Independently of the plans for nuclear power at national level, every operating nuclear power plant (NPP) will eventually close and enter the decommissioning phase. In the European Union, 33 regions host some operational NPP with closure scheduled before 2050. This paper reviews the economic- and human capital-related challenges for local communities by means of various case studies of NPP closures.
The study is carried out via a desktop research of publicly available data. The first observation is that the regional and local socio-economic impacts are not at the top of the legislative agenda. National policies typically set the focus on public health and safety, energy supply, environmental impacts and technical feasibility. However, the socio-economic impacts have begun to gain greater importance with the increasing rate of NPP closures. To a certain extent, the experience is similar to any other industrial transition perceived from the local perspective. Commonly, the operator of the NPP is a large employer contributing to the economic prosperity of the region often located in a rural area. Nevertheless, our preliminary results show specificities of the nuclear sector that need to be taken in to account in the design of recovery plans.
NPPs bring high-skilled and high-paid workers and their families into relatively small towns. NPPs provide jobs, not only related to their operation, but also in sub-contractors and indirectly associated activities. This has created a sense of community pride keeping the local economy vivid. In addition, the community around nuclear facilities provide a substantial income to the region from property taxes and other revenues. Another specific factor is the presence of spent nuclear fuel at decommissioned sites, which may represent an obstacle to the site redevelopment. In a short-term, new opportunities will rise in the vicinity rather than directly on the site.
Once a plant announces its closure, it is important to identify the possible impacts and estimate changes in the number and structure of affected jobs. A sufficient number of employees is required to ensure safe plant operation through the processes of decommissioning and dismantling. The age structure of the staff is also important. More aged employees may opt for early retirement. Younger generations, in general, have a larger potential for requalification. Workers with transferrable skills may be less affected and seek local re-employment. The regions should revise potentials for economic growth covering different opportunities and attract new investors matching wishes and skills of the employees. The proposal of concrete revitalization plans will provide incentives retaining the qualified workers in the region.
The purpose of this study is to analyze the different lessons learned and show best practices across the Europe. Our review shows that each host community has its own experience and challenges. However, as shown in a few examples above, we were able to derive some common factors which if considered in the coping strategy developed well before NPP closure may reduce the risks of job losses and a corresponding reduction in tax revenues in the affected regions.

09.09.2021 15:00 Education, training and outreach

Education, training and outreach - 1103

Towards Optimized Use of Research Reactors in Europe - the TOURR Project

Pavel Gabriel Lazaro

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

gabriel.pavel@enen.eu

The primary objective of the TOURR project is to develop a strategy for Research Reactors (RRs) in Europe and prepare the ground for its implementation. This strategic goal can be divided into specific objectives: Assessment of the current status of European research reactors fleet, including plans for upgrades, evaluation of urgent EU needs, developing tools for optimal use of the research reactors fleet and finally, rising awareness among decision-makers on the (future) role of research reactors.
The ambition of the TOURR project is to secure access and availability of RRs as a vital part of the European Research Area and support a stable supply of medical radioisotopes. Nuclear RRs have been constructed in countries implementing nuclear power plants and used in experiments necessary to develop commercial reactors and training programmes.
Neutron irradiation has found new applications in the adaption and production of existing and new materials, including medical radioisotopes. The latter enabled the development of new diagnosis and treatment techniques, for the benefit of millions of patients.
Europe has a broad and very diverse landscape of research reactors, many of them have already been for 30-50 years in operation, well maintained and regularly upgraded. Yet financial pressure, caused by a combination of declining interest and the absence of a sound financial model, led to the closure of many of them. On the other hand, only one research reactor is now being constructed – the Jules Horowitz Reactor, in Cadarache.
Those negative trends call for a coordinated European action to assess the impact of the decreasing number of research reactors, identify future needs (including new neutron sources), draw a roadmap for the upgrade of the existing research reactors fleet, and a model for harmonized resource management. The TOURR project is a response to this challenge.
The paper will introduce the TOURR project as a whole, focusing on the main envisioned outcomes and the methodology that will be used to achieve them. The project is divided in to 5 work packaged which will be presented in detail. The paper will serve as an overview and reference of all future TOURR publications.

ACKNOWLEDGMENTS
This project receives funding from the EURATOM Research and Training programme 3 years under grant agreement N° 945269.

TOURR CONTRIBUTORS
European Nuclear Education Network, ENEN, Belgium: Gabriel Lazaro Pavel, Roberta Cirillo, Francisco Suárez Ortiz
Centrum Výzkumu Řež, CVR, Czech Republic:
Energiatudományi Kutatóközpont, EK, Hungary: László Szentmiklósi, Péter Juhász
Narodowe Centrum Badań Jądrowych, NCBJ, Poland: Renata Mikołajczak, Jacek Gajewski, Grzegorz Krzysztoszek, Iliana Chwalińska, Małgorzata Kot, Joanna Walkiewicz
Belgian Nuclear Research Centre SCK CEN: Michele Coeck, Lisanne Van Puyvelde Institut “Jožef Stefan”, JSI, Slovenia: Luka Snoj, Bor Kos, Anže Pungerčič, Vladimir Radulović, Anže Jazbec, Saša Škof
Evalion SRO, EVALION, Czech Republic: Petr Koran, Michaela Velckova
Universität Stuttgart, USTUTT, Germany: Georg Pohlner, Joerg Starflinger
Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, CIEMAT, Spain: Daniel Cano, Enrique Gonzalez

09.09.2021 15:20 Environment and back end of the fuel cycle II

Environment and back end of the fuel cycle - 1006

Method for analysis of neutron activation measurements of Am-241 with uncertainty propagation

Gašper Žerovnik1, Vladimir Radulović1, Ljudmila Benedik2, Bor Kos1, Tjaž Gantar3, Gilles Noguere4

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

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

3Fakulteta za matematiko in fiziko, Jadranska 19, 1111 LJUBLJANA, Slovenia

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

gasper.zerovnik@ijs.si

Apart from being used for numerous applications, Am-241 is an important component of spent nuclear fuel due to its contribution to the decay heat. Due to its inherent ?-ray emission, accurate time-of-flight (TOF) measurements of capture cross section in Am-241 are difficult to perform. Therefore, there is a potential to obtain a more accurate value of the cross section for thermal neutron capture in Am-241 using neutron activation measurements. This value may eventually serve for normalisation of the energy dependent capture yields measured by TOF. Neutron activation analysis of Am-241 is comparatively complex for several reasons. First, the Am-241 cross section contains a resonance below the Cd transmission filter cut-off energy (~0.55 eV) and another one overlapping with that energy. Second, whereas it is not unique that the activation product is produced in both ground and metastable state, it is rare that the latter has a much longer half-life (141 a for Am-242m vs. 16.02 h for Am-242g). And finally, the decay scheme of the activation products is relatively complicated and difficult to measure. ?-ray spectrometry of activation products is extremely difficult due to low ?-ray energies and increased ?-ray background from e.g. Am-241 fission products or possible activation products in substrate or container materials which are produced by neutron irradiation.

Therefore, the only realistic option is ?-particle spectrometry, originating from Cm-242, which is a decay product of Am-242g,m. Due to the much longer half-life of the metastable product and a relatively small branching fraction for its production from neutron capture (~0.09), its contribution to the Cm-242 activity is very low in absolute terms and negligible compred to the contribution from the ground state for a few years after irradiation. In order to more accurately determine the neutron fluence, at least two standard monitor materials need to be irradiated simultaneously with Am-241. In this case, Co-59(n,?), Au-197(n,?) and Ni-58(n,p) reactions were chosen for this purpose. Each measurement set consists of two irradiations, one with and one without cadmium transmission filter. The activities of samples of monitor materials are measured by a ?-ray spectrometer.

The main goal of this paper is to outline the major sources of uncertainty and develop a rigorous method for uncertainty propagation from measured count rates to corresponding reaction rates, properly taking into account all independent and common sources of uncertainty. In the past, many of these were treated as either independent of fully correlated. The end result is a vector of reaction rates (for all irradiated materials with and without Cd) with the corresponding full covariance matrix.

The next step, the calculation of the Am-241 thermal capture cross section from the reaction rates, obtained here, is the final aim of the project. If done successfully, the final result may prove useful for future nuclear data evaluations.

09.09.2021 15:40 Environment and back end of the fuel cycle II

Environment and back end of the fuel cycle - 1021

Fukushima nuclear accident impacts on the EU nuclear arena

Helena Janžekovič

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

helena.janzekovic@gov.si

The Fukushima Daiichi nuclear power plant (NPP) accident happened ten years ago in Japan. It was the very first severe nuclear accident initiated by natural events, namely tsunami which followed the earthquake. Although geographically very far from Europe the accident has a profound impact on the nuclear arena in the European Union (EU) as a whole in addition to its impact on particular Member State (MS) nuclear frameworks. Moreover, it might have the influences also in the next decades.
The immediate consequences were related to protection of EU citizens in Japan and in the region affected as well as to trafficking from Japan, assuring that no contaminated goods, airplanes and ships were either imported in EU or landed in EU MSs as appropriate. Among such goods special attention was given to the contamination of food, i.e. either food from Japan or from fishing areas affected by atmospheric and liquid releases from the Fukushima Daiichi site in the Pacific Ocean. The EU legislation related to the control of food required regular updating or amending. In very first days of the accident not only trafficking from but also trafficking to Japan was heavily affected.
The footprint of the accident was also related to the contamination of the environment in particular to the atmospheric releases which took place for weeks and were eventually also measured in Europe.
Regarding the NPPs and other nuclear facilities the so-called stress tests took place in the EU MSs followed by the upgrading of NPPs and other nuclear facilities in order to prevent severe accidents and to mitigate their consequences. These upgrading were related to substantial investments. Moreover, in some MSs the attitude related to the operation of NPPs changed. A vivid discussion on nuclear energy is going on based on the Technical assessment of nuclear energy with respect to the ‘do no significant harm’ criteria of Regulation (EU) 2020/852 (‘Taxonomy Regulation’) prepared by the JRC in 2020.
The EU MSs also updated several legal acts in order to become more effective in prevention of such accidents on one site while assuring adequate response when such accident happen. Among such legal act is the Nuclear Safety Directive which has been amended in 2014 leading to several additional activities exercised in the EU, such as peer reviews. Moreover, the accident also influenced the drafting other legal acts such as basic safety standards for protection against the dangers arising from exposure to ionising radiation which has been published as Council Directive 2013/59/Euratom in 2013.
In addition, the cooperation among the EU MSs and the IAEA strengthen. Regarding emergency preparedness the HERCA/WENRA approach addressed zoning around NPPs to be a subject of measures in case of an emergency. The holistic approach to such measures has be underlined in order to protect the most vulnerable members of the public in such zones. Development of so-called post-accident strategies started and the term “citizen-science” in nuclear arena has been applied. On the other hand, several research programs within HORIZON 2020 and Horizon Europe were initiated, e.g. interaction of melted core with other materials and development of drones to be used during or after severe accident. These are only some of the changed and activities within the EU MSs identified.
The article given a systematic overview of changes which occur in the radiation and nuclear framework in EU MSs in a decade with followed the Fukushima Daiichi nuclear accident.

09.09.2021 16:00 Environment and back end of the fuel cycle II

Environment and back end of the fuel cycle - 1005

Approach for the Prospective Assessment of Radiological Impact to the Population from Decommissioning of “RADON” Type Radioactive Waste Storage Facility in Lithuania

Valdas Ragaišis, Povilas Poskas, Audrius Šimonis

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

valdas.ragaisis@lei.lt

The Maišiagala radioactive waste storage facility in the Republic of Lithuania is located approximately 30 km to North-West from the capital city Vilnius. The facility is of a former Soviet Union “Radon” type facility design and initially was planned for the disposal of institutional waste. The facility started operation in 1963. Disused spent sealed sources and solid waste from industry, military and medicine applications from Lithuania and neighboring countries were disposed in bulk, in one sub-surface reinforced concrete vault of 200 m3 volume. The facility was closed in 1989 being partially filled with the waste. Concrete layers and sand occupy the residual volume. The waste is stored in various forms, inventory is estimated to be 7.7E+13 Bq. The most activity contributing radionuclides are short-lived H-3 and Cs-137. Other significant radionuclides are long-lived Pu-239, C-14, Ra-226 together with short-lived Sr-90, Co-60.
The Maišiagala facility cannot be upgraded to a disposal facility which would meet the modern environmental and radiation safety standards. Therefore, a decision was taken to retrieve the waste, pre-sort and transport to the Ignalina NPP site, where the waste could be appropriately characterized, sorted and disposed or interim stored until appropriate disposal facility will become available in Lithuania. The current operator of the facility, Ignalina NPP, is responsible for the implementation of decommissioning activities.
The waste retrieval concept foresees the construction of a light, air-tight structure, which will enclose the waste storage vault and where the waste retrieval, pre-sorting and packaging equipment will be installed and operated. The work on waste retrieval, packaging and off-site shipment would be carried out during the warm season of the year, presumable from April to November. Two – three seasons would be required for the decommissioning of the site to the “green field” conditions.
The emission of airborne radionuclides to the environment and exposure of the population during the performance of decommissioning activities are estimated. The emission is expected to be irregular and will vary during the year, week, day and the performed activity. The knowledge on the stored waste inventory is rather uncertain and cautious assumptions are made, especially regarding emission of gaseous Rn-222. The presentation discusses the approach for the assessment of expected releases of radionuclides to the environment and the approach for the consideration of radiological impact to the population.
Increase of external irradiation fields in the close proximity to the site and during the waste transport will create additional exposure pathways. Assessment of direct irradiation is rather straightforward and also will be discussed in the context of its significance to the overall impact to the population from the planned activity.

08.09.2021 11:00 NARSIS special session

NARSIS special session - 1201

Study of the Seismic Behaviour of Gen III Plant: The Influence of Ageing

Rosa Lo Frano1, Salvatore Angelo Cancemi1, Pierre Gehl2, Evelyne Foerster3

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

2BRGM, 3, Avenue Claude Guilleunia, BP 36009, 45060 Orleans, France

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

rosa.lofrano@ing.unipi.it

External events are a significant source of hazards to nuclear power plant operation, and for this reason it is of meaningful importance to investigate the existing operational nuclear reactor, even aged, in order to verify their structural capacity adequacy.
This study aims at investigating numerically, by means of a deterministic approach, the dynamic behaviour of the structures, systems and components (SSCs) relevant for the safety of the plant when subjected to earthquake event.
A quite refined FEM model was thus set up and implemented considering suitable aged/not aged materials behaviour and constitutive laws for the SSCs material. Damping ratios of building materials of H>50 m (reinforced concrete and steel) have been adopted as well.
Fifty acceleration time histories, representative of different soil conditions, were inputted to the transient analyses.
The obtained results were used to appropriately check mainly the considered NPP containment strength reserve.
Results seemed to confirm the overall containment reliability even buckling could affect some internal components. It was observed also that aging determines a reduction of the plant structural capacity of about 20%.

08.09.2021 11:20 NARSIS special session

NARSIS special session - 1202

Multiple Hazard Modeling Utilizing Traditional PSA Tools

Aleksej Kaszko1, Slawomir Potempski2

1National Centre for Nuclear Research, ul. Andrzeja Sołtana 7, Otwock-Świerk, Poland

2Narodowe Centrum Badań Jądrowych, Soltana 7, 05-400 Otwock, Poland

aleksej.kaszko@ncbj.gov.pl

External multi-hazard Probabilistic Safety Assessment (PSA) for more than a decade is a significant concern for nuclear scientists worldwide. Therefore one of the objective of the NARSIS project was to propose a practical method for multiple-hazard probabilistic safety assessment. This method utilizes current PSA software capabilities with knowledge received from Probabilistic Hazard Assessments. It mainly utilizes PSA models that every nuclear facility already has; therefore, its implementation costs should be bearable by every facility.
This paper presents a methodology and an example case with earthquake and flooding hazards during the Loss of Off-site Power scenario. A similar accident happened in Fukushima, therefore it can be a good case for evaluation of the proposed methodology. This example shows that although multiple hazards are rare, they can pose a significant danger if no precautions are taken into account.

08.09.2021 11:40 NARSIS special session

NARSIS special session - 1203

Seismic fragility analysis based on vector-valued intensity measures; theory and application to fuel assembly grids

Manuel Pellissetti1, Marie-Cécile Robin-Boudaoud2, Pierre Gehl3

1Framatome GmbH, Paul-Gossen-Str. 100, 91052 Erlangen, Germany

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

3BRGM, 3, Avenue Claude Guilleunia, BP 36009, 45060 Orleans, France

manuel.pellissetti@framatome.com

Standard practice in seismic fragility analysis is to rely on a single (scalar) intensity measure (IM), typically horizontal peak ground acceleration (PGA).
The present paper discusses the benefits of using vector-based IMs (or “vector-IMs”). A case study dealing with the seismic fragility of fuel assembly spacer grids is presented. The uncertainties induced by scalar-IM fragility curves or vector-IM fragility functions are compared and discussed.
Globally, a reduction in the dispersion is observed. However, care should be taken when interpreting vector-IM fragility functions that are based on strongly correlated ground-motion parameters.

08.09.2021 12:00 NARSIS special session

NARSIS special session - 1204

SEVERA Decision Support Tool Development in Project NARSIS

Luka Štrubelj1, Marko Bohanec2, Ivan Vrbanić3, Ivica Bašić3, Klemen Debelak1

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

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

3APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia

luka.strubelj@gen-energija.si

The project NARSIS – New Approach to Reactor Safety ImprovementS is making scientific step towards addressing the update of some elements required for the safety assessment. These improvements mainly concern:
• Natural hazards characterization, in particular by considering concomitant external events, either simultaneous-yet-independent hazards or cascading events, and the correlation in intra-event intensity parameters.
• Vulnerability of the elements to complex aggressions, with the integration of new approaches such as vector-based fragility surfaces and reduced models
• Better treatment of uncertainties through adoption of probabilistic framework for vulnerability curves and non-probabilistic approach to constraining the “expert judgments”.
• Develop decision support tool for severe accidents
In first step the referential NPP is identified: pressurized water reactor NPP with two loops. In second step the severe accident management guidelines for referential NPP are characterized. In third step the relevant scenarios are identified, hazard-induced damage states defined and state-specific accident progression event tree for demonstration purposes are developed. In fourth step the applicable deterministic analyses of severe accidents are performed. In last-fifth step the decision support tool for severe accidents – SEVERA is developed. Its purpose is a prototype demonstration-level decision support system aimed at supporting the technical support center while managing a severe accident. The SEVERA represent, store and monitor selected physical measurements of the NPP. It assesses the current state of barriers: core, reactor coolant system, reactor pressure vessel and containment. The prediction of future accident progression, if no action is undertaken is one of basic functions. The support tool provides a list of possible management recovery strategies and courses of action. The applicability and feasibility of possible actions in the given situation is identified. For each action the prediction of the consequences in terms of probability of the last barrier (containment) failure and estimated time window for failure. At the end the SEVERA evaluate and rank the feasible actions, providing recommendations for the TSC.

07.09.2021 10:10 Poster session RED

Advances in nuclear technology - 204

Dynamic response of LFR in cogeneration mode

Riccardo Chebac1, Marco E. Ricotti2, Antonio Cammi3, Khashayar Sadeghi4, Seyed Hadi Ghazaie4, Ekaterina Sokolova4, Evgeniy Fedorovich4

1Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy

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

3Politecnico di Milano, Department of Energy, Via La Masa 34, 20156 Milano, Italy

4Peter the Great St. Petersburg Polytechnic University, 195251 Russian Federation

riccardo.chebac@polimi.it

In this paper a novel object-oriented simulator capable of transient analysis of the Advanced Lead-cooled Fast Reactor European Design, based on the model envisioned by Ponciroli et al.[1], was developed within the Dymola software based on the Modelica language. The nuclear power plant (NPP) model was conceived as part of a more complete simulator capable of studying the technical feasibility of hybrid systems scenarios. The plant was built in such a way to enable steam extraction from the turbine unit therefore enabling the user to study different cogeneration options. The reactor core behaviour was implemented via point reactor kinetics coherently with ALFRED specifications. For the Steam Generator (SG) a single block modelling the effect of eight bayonet-type SGs was developed. Temperature of the extracted steam from the low-pressure turbine (LPT) can be determined by the sink pressure, which makes the model more flexible in studying the various cogeneration systems. Three scenarios were studied, namely: control rods step insertion and extraction, water mass flowrate linear increase and extraction valve regulation. Results highlight the time constants of the various components and show the potentials of steam extraction which will indeed require further investigation and development of a suitable controller in order to efficiently use this system.

07.09.2021 10:10 Poster session RED

Advances in nuclear technology - 207

The potential of the TEPLATOR application in the Central Europe region

David Mašata1, Radek Škoda2, Lucie Noháčová1

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

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

masata@kee.zcu.cz

The innovative emissions-free concept of future district heating source TEPLATOR was introduced in 2020 by a group of researchers from CTU in Prague and UWB in Pilsen in the Czech Republic. TEPLATOR will use already spent nuclear fuel assemblies for a clean and economical heat-only production with an output range between 50 MWt and 150 MWt. This article evaluates the progress of the TEPLATOR applicability potential development, especially focusing on the TEPLATOR fuelling.

The TEPLATOR facility potential location assumes a significant heat demand exceeding 1,5 PJ per year and a currently operating district heating network is required. Then the production price of heat generated by TEPLATOR could reach less than 2 EUR per GJ of energy, which indicates an attractive alternative to the outdated coal-based heating plants and new intended gas boilers. The popularity of district heating varies across Europe and there currently exist many promising locations for the TEPLATOR application.

The fuelling of TEPLATOR will be provided with the spent fuel from NPPs operating VVER, PWR, or BWR. The overview about the reserves of available fuel assemblies in EU countries is presented and assessed for the TEPLATOR use. The possibilities of spent nuclear fuel transportation have been investigated and evaluated. The final economic assessment of spent fuel utilization is provided. An alternative approach for the countries with no light water reactors in operation represents slightly enriched fresh nuclear fuel. The economic study of the TEPLATOR operation with fresh fuel has been performed and compared with the possibility of spent nuclear fuel use.

07.09.2021 10:10 Poster session RED

Advances in nuclear technology - 210

Evaluating the TEPLATOR Concept as a Nuclear Heating Solution Compared with other Alternatives in the Czech Republic.

Hussein Abdulkareem Saleh Abushamah1, Jana Jiřičková2, Radek Skoda2, David Masata2

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

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

abushama@fel.zcu.cz

In the Czech Republic, heating sector accounts for about 56% of the total final energy consumption. District Heating Systems (DHSs) supply nearly 40% of the residential heating demand where the heat sources are mainly coal-based heat-only or CHP plants, causing significant emissions of CO2. For limiting CO2 emissions, the termination of coal combusting in the Czech Republic is expected to be in 2038. Consequently, the energy policy makers and the companies are intensively looking for alternatives to fossil fuel-based plants for serving the heating needs. The alternatives are categorized into two approaches namely electric based and non-electric based DHSs where different kinds of primary energy sources contribute to electricity generation or to only-heat generation. In both approaches nuclear energy can be one of the primary energy sources. On the other side, a nuclear heat-only concept namely TEPLATOR is under development to be a carbon-free solution for the heating sector. In this study TEPLATOR concept is introduced, the heating sector and the available energy sources in the Czech Republic are discussed. Moreover, the advantages and disadvantages of TEPLATOR heating concept compared with the other alternative options such as coal, natural gas, biomass, and electric based heating technologies, are presented. Carbon emissions reduction, lower heat generation cost, economizing the expansion of DHSs in the lower heat demand density areas are some of the important advantages which are evaluated in this study. Finally, the benefits of this approach are formulated and calculated for a typical case study in the Czech Republic which confirm the superiority of this nuclear heat-only concept in district heating applications.

07.09.2021 10:10 Poster session RED

Reactor physics & research reactors - 308

Further development of RAPID code extension for TRIGA reactor 3D burnup calculations

Anze Pungercic1, Alireza Haghighat2, Luka Snoj1

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

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

anze.pungercic@ijs.si

Determination of accurate 3D pin-wise fuel burnup in nuclear reactors is essential for 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 entire operational history that allows for burnup validation possibilities in the form of measured excess reactivity.
In the paper, we describe the extension of the bRAPID algorithm for its application to the TRIGA reactor. In particular, we will focus on bRAPID’s database pre-calculation procedure in which the Fission Matrix (FM) coefficients for different combination of reactor power and irradiation times are calculated. The FM coefficients are dependent on fuel burnup and change with U235 depletion, Pu239 production and reactor poison (xenon and samarium) formation. The evaluation of such changes is crucial for the development of the bRAPID burnup methodology.

07.09.2021 10:10 Poster session RED

Reactor physics & research reactors - 311

Investigation of Recently Introduced Diffusion Coefficient Generation Methods

András Szabolcs Ványi1, Mathieu Hursin2, Szabolcs Czifrus1

1Institute of Nuclear Techniques Budapest University of Technology and Economics, Muegyetem rkp. 9, H-1111 Budapest, Hungary

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

vanyi.andras@reak.bme.hu

The multigroup diffusion theory is one of the most widely used methods for deterministic reactor core calculations. In this approach the angular dependence of the neutron flux and the scattering kernel is expanded with spherical harmonics to the P$_1$ order. The P$_1$ scattering matrix is then used to generate a scalar quantity for each energy group, the group diffusion coefficient. As the entire linearly anisotropic angular dependence is represented in the group diffusion coefficients, the accuracy of the diffusion calculations highly depends on how those coefficients are determined. Recently, two approaches were introduced by researchers of the Massachusetts Institute of Technology to produce accurate diffusion coefficients using Monte Carlo codes: the Neutron Leakage Correction method~\cite{Herman_2013} and the Cumulative Migration Method~\cite{Liu_2016}. After providing a brief overview of the methods, in this paper a parametric study is carried out where the performance of those approaches in terms of power distribution and multiplication factor of the associated diffusion calculations is assessed. For this purpose, 1D and 2D models based on the specifications of the DIMPLE benchmark \cite{DIMPLE} are used. The Monte Carlo simulations for group constant generation and reference calculations were carried out by the Serpent 2 code, while the deterministic calculations were performed by the PARCS nodal diffusion code.

\begin{thebibliography}{99}
\bibitem{Herman_2013}
B.R. Herman, B. Forget, K. Smith, B. N. Aviles, Improved diffusion coefficients generated from Monte
Carlo codes.'',
Technical report, American Nuclear Society, 555 North Kensington
Avenue, La Grange Park, IL 60526 (United States), 2013

\bibitem{Liu_2016}
Z. Liu, K. Smith, B. Forget, A Cumulative Migration Method for Computing Rigorous Transport Cross Sections and Diffusion Coefficients for LWR Lattices with Monte Carlo'',
PHYSOR, pp. 2915–2930, 2016

\bibitem{DIMPLE}
D. Hanlon, S. Assurancet, Light Water Moderated and
Reflected Low Enriched Uranium (3 wt.\% $^{235}$U) Dioxide Rod
Lattices'',
NEA/NSC/DOC(95)03/IV, Vol.~4, LEU-COMP-THERM-055, 2002
\end{thebibliography}

07.09.2021 10:10 Poster session RED

Reactor physics & research reactors - 314

Jožef Stefan Institute TRIGA Research Reactor Activities in the Period from September 2020 – August 2021

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

anze.jazbec@ijs.si

The Jožef Stefan Institute (JSI) has been operating a 250 kW TRIGA research reactor since 1966. Safety performance indicators (SPI) have been monitored for over 10 years. Examples of the monitored parameters are the operating time, the number of irradiated samples, doses received by operating staff, and the activity of radioactive gases released to the environment. In the paper, SPIs for the year 2020 will be presented and analyzed. Also, a comparison among SPIs from previous years will be made. In the paper, it will be presented how the virus SARV-CoV-2 affected the operation in the last year.
Furthermore, new research work carried out during the years 2020 and 2021 will be presented. Some research campaigns continued from the previous years, like the NATO-SPS project e-SiCure2 and collaboration with industrial partners, e.g. Rolls Royce.
In spring 2021, an extensive campaign was performed aimed at the development of a reactor power meter based on a Cherenkov radiation detector, applicable for both steady-state and pulse operation modes. The measurement set-up was developed and tested within a Master thesis project at the Reactor Physics department at JSI.
In the same period, test irradiations were performed in collaboration with the National Institute of Chemistry. The purpose of the project is to transform various waste chemicals containing hydrogen into methanol using radiolysis. Special attention was given to various catalysts which could increase methanol production.
Due to the COVID pandemic situation, all education and training activities were performed online. Some experience was gained during the training of future NPP staff in summer 2020. The first fully remote course was performed for students from the Uppsala University, Sweden, in September 2020. Due to the good experience, all practical exercises for Slovenian students in autumn 2020 were organized online, with excellent feedback overall. Knowledge and practical experience in remote education activities will be of great value in the context of the EC-funded project ENEEP (European Nuclear Experimental Educational Platform - http://eneep.org/), facilitating access to practical education activities to a wider user base.
In the last year, all visits to the reactor were canceled. However, this did not mean that the general public could not visit us: the operating team participated in two virtual events, the first one is called Noč Raziskovalcev – Researchers' Night, and the second one was JSI Open Day. Visitors joined us at Zoom events and together, we did a virtual walk through the facility. Besides, everybody could ask questions which were later addressed by the presenter.
Since the reactor operated less than in previous years, more activities were focused on facility maintenance. We performed an upgrade of the physical protection system and a renovation of the reactor safety system. There is a plan to modernize the water purification system inside the spent fuel pool.

07.09.2021 10:10 Poster session RED

Reactor physics & research reactors - 317

Analyses of Doppler Coefficient of Reactivity using Monte Carlo and Deterministic codes

Dušan Čalič

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

dusan.calic@ijs.si

Using Mosteller pin cell benchmark the Doppler coefficient for UO2 fuel was analysed using Monte Carlo code Serpent and deterministic codes Dragon and WIMS. These coefficients were calculated using nuclear data taken from ENDF/B-VI.8, ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0. Additional for the deterministic codes the analysis was performed using different energy structure. This paper presents the results from the listed codes and compares their results.

07.09.2021 10:10 Poster session RED

Reactor physics & research reactors - 320

Monte Carlo simulation of k0 instrumental neutron activation analysis

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

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

lojze.gacnik@gmail.com

Neutron activation analysis (NAA) is a powerful set of methods for elemental composition analysis. However, the standard approach used to calculate the elemental composition from NAA measurements has many limitations. We replaced the current method of data analysis with computer simulation. By using the Geant4 particle simulation toolkit, we simulated NAA from the initial neutrons that activate the sample, to the final gamma rays hitting the high purity germanium (HPGe) detector.
To compare simulation results with measurement, we used a simple model of detector electronics, that converts energy deposition events within the detector, to the energy recorded by the electronics, by taking into account both true and random coincidence effects. NAA thus became an optimization problem - to minimize the difference between simulation output and experiment, by finding the appropriate simulation inputs, e.g. the sample composition, or HPGe detector geometry. This unified the up to now disparate parts of k0 instrumental neutron activation analysis (k0-INAA) - sample composition determination, detector calibration, and neutron spectrum determination all reduce to the same optimization problem. The only difference between them is which parameters are known versus unknown. This simplifies activation analysis, and improves accuracy, as far fewer approximations are made.
We tested this approach by comparing its results versus traditional k0-INAA on certified reference materials.

07.09.2021 10:10 Poster session RED

Severe accidents - 407

Numerical simulations of prototypical oxide-metal corium heat by electromagnetic induction

Julien Guillou1, Philippe Tordjeman2, Wladimir Bergez3, Rémi Zamansky3, Jean-Francois Haquet1, Pascal Piluso4, Anne Boulin1, Sebastien Renaudiere De Vaux3

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

2Electricite de France (EDF) R&D, Mecanique des Fluides Energie et Environnement, 6 quai Watier, 78401 Chatou, France

3The Institute of Fluid Mechanics, Toulouse, 2 Allée du Pr Camille Soula, 31400 Toulouse, France

4CEA-Cadarache, Severe Accident Experimental Laboratory Bd 219, 13 108 Saint-Paul-Lez-Durance, France

julien.guillou@imft.fr

In the frame of a hypothetical nuclear accident leading to the partial or the complete melting of the nuclear fuel, the core could melt and create a high-temperature mixture called corium. Accordingly, the mixture could break through the vessel, spread over the basemat of the reactor cavity, and interact with the concrete of the reactor pit, described in the literature as Molten Core Concrete Interaction (MCCI). This interaction results in the concrete ablation due to the high heat flux from the corium (150 kW/m2). To understand this phenomenon, experiments using prototypical corium were performed at CEA Cadarache PLINIUS platform within VULCANO facility and using induction heating to simulate the residual decay heat coming from the fission products within the corium.
In case of an oxide-metal corium, the volumetric power is uniform and mainly within the melt oxide phase (98%). It has to be checked that the induction heating is representative of this residual power partition. Similarly, the Lorentz force induced by the selected heating technology may change the melt hydrodynamic and could affect the convective heat transfer coefficient with the concrete. In this study, the corium is composed with 90% of oxide and 10% of metal.
Firstly, our study investigates the ability of an induction process to perform an uniform Joule heating and the same repartition when the melt is immersed in an AC magnetic field generated by coils, simulating reactor case configuration.
Induced current mainly occurs inside the electromagnetic skin depth, which is linked to the electrical conductivity of the considered phase and the frequency of the harmonic current inside the coil.
Preliminary, in order to determine the AC field characteristics (frequency and intensity) and the metallic dispersed phase size and its spatial scattering corresponding to the residual power distribution, 3D simulations are performed under COMSOL considering a cylindrical vessel filled with a solid oxide and metallic balls inside a coil. Design case is reached when all of them are closed to the outer walls inside the electromagnetic skin depth.
Secondly, in order to study the Lorentz force influence on the convective structures and the heat transfer coefficient with the concrete, magneto-hydrodynamic simulations are conducted from this design case with JADIM code developed and validated at IMFT institute. This part focuses on metallic drops immersed in a non-electrical conducting fluid exposed to an AC magnetic fluid (assuming that the metal conductivity is 1000 times more effective than the oxide one).
These simulations should increase the understanding of the corium melt thermo-hydraulics under induction heating and its ability to be used in the scope of prototypical corium experimental devices as being representative of the decay heat.

07.09.2021 10:10 Poster session RED

Severe accidents - 408

Investigation on Revaporization from CsI Deposited Particles in the Primary Circuit in Nuclear Severe Accident Conditions

Melany Gouello1, Teemu Kärkelä2

1VTT, Tietotie 3, FI-02150 Espoo, Finland

2VTT Technical Research Centre of Finland, Tietotie 3, Espoo, 02044 VTT, Finland

melany.gouello@vtt.fi

Since the Fukushima Dai-Ichi (1F) Nuclear Power Station accident, a renewed attention has been posed on the evaluation of delayed source terms, especially from Reactor Coolant System (RCS) surface deposits and on the boron carbide control rods behaviour. After fission products have been released from the overheated and molten fuel, they are transported through the reactor coolant system and fission products will reach areas at lower temperature. Therefore, vapour condensation and particle nucleation processes take place in the gas flow. If vapour condensation takes place close to the surfaces of the primary circuit, a layer of condensate can be formed on it. Particles in the gas flow may also deposit on the circuit surfaces together with control rod and structural materials.
The objectives of the work was to investigate the behavior of deposited caesium iodide particles in the RCS and to assess the effect of gaseous boric acid with the deposited particles.
Caesium iodide particles were generated by nebulization of a concentrated caesium iodide aqueous solution and then passed through a Thermal Gradient Tube (TGT). The gaseous boric acid was generated by vaporization from a crucible placed inside the furnace. Aerosols and gaseous species were sampled at 150°C on filters and liquid scrubbers and analyzed with HR-ICP-MS. Aerosol number size distributions were measured with TSI Scanning Mobility Particle Sizer (SMPS), with series 3080 platform, series 3081 Differential Mobility Analyzer (DMA) and series 3775 Condensation Particle Counter (CPC). The aerosol mass concentration was monitored by Tapered Element Oscillating Microbalance Series 1400a (TEOM).
The study showed that the deposited caesium iodide particles were subject to revaporization process in Ar/H2 atmosphere; gaseous iodine was released from the deposits. When gaseous boric acid was present in the carrier gas (Ar/H2), the percentage of gaseous iodine released was higher. The interpretation of the SMPS and TEOM results also suggested that the presence of gaseous boric acid would then have an influence on the transport of the deposited caesium iodide particles as it showed different behaviour than during revaporisation/resuspension under Ar/H2.

07.09.2021 10:10 Poster session RED

Severe accidents - 411

Analysis of stratified steam explosion in reactor conditions

Matjaž Leskovar, Janez Kokalj, Mitja Uršič

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

matjaz.leskovar@ijs.si

Steam explosions are one of the most feared potential hazards that may occur in the course of a severe accident in a nuclear power plant with core meltdown. It can threaten the integrity of the reactor containment and lead to early release of radioactivity. A steam explosion is an energetic fuel coolant interaction process, which may occur when the hot reactor core melt comes into contact with the coolant water. Recently an important safety related uncertainty was revealed, which is related to the experimental observation of unexpected strong spontaneous steam explosions in stratified melt-coolant configurations.

One of the important conditions for the possible occurrence of an energetic steam explosion is the formation of an appropriate premixture of pre-fragmented melt and coolant. Stratified melt-coolant configurations, i.e. a molten corium layer below a coolant layer, were believed as being unable to generate strong explosive interactions because of the limited amount of existing premixture in favourable conditions for steam explosion. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with high melting temperature oxidic simulants of corium revealed that strong steam explosions may develop spontaneously in stratified melt-coolant configurations. In the tests the formation of a considerable melt-coolant premixed layer was observed prior to the explosions. The observed strong spontaneous stratified steam explosions have an important impact on the safety related issue of fuel coolant interactions in nuclear power plants.

In the paper, the study of stratified steam explosions in reactor conditions performed with the MC3D code (IRSN, France) will be presented. A stratified steam explosion model was developed and validated on available PULiMS and SES experimental data. To get a better insight into stratified steam explosions in reactor conditions a comprehensive parametric analysis was performed, varying the premixed layer thickness, fraction of phases in premixture, water layer thickness, melt spreading area and triggering location. The calculations were compared to the previously performed study of steam explosion in ordinary melt jet-coolant pool configuration. Various simulation results will be presented and discussed. Suggestions for further analytical and experimental work will also be given.

07.09.2021 10:10 Poster session RED

Severe accidents - 414

Simulation of Experiment on Passive Autocatalytic Recombiner Phenomena

Tilen Švarc1, Jure Marn2, Ivo Kljenak3

1Univerza v Mariboru, Smetanova ulica 17, pp 224, 2000 Maribor, Slovenia

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

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

ivo.kljenak@ijs.si

During a severe accident in a light water reactor nuclear power plant, large amounts of hydrogen could be generated due to the oxidation of the reactor core. The formation of regions in the containment with high local hydrogen concentration could increase the risk of hydrogen explosion, which could in turn threaten the containment integrity. Among other possible mitigation systems, Passive Autocatalytic Recombiners (PARs) are being used to reduce the hydrogen quantity in the containment. These are box-shaped devices through which gas from the containment atmosphere flows upwards, with hydrogen recombining with oxygen from the air on catalyst-coated vertical plates. The flow occurs as natural convection driven by the generated reaction heat.

Experiments are being performed in experimental facilities that suitably replicate PARs, with the purpose of a better understanding of their functioning and further development. Simulations of phenomena that occur in these facilities, applying description on the local instantaneous scale (by using so-called Computational Fluid Dynamics – CFD – codes), first enable the validation of relevant physical models. After that, the analyses of simulation results provide additional insights into the alleged local behavior of physical variables: velocity, temperature and hydrogen concentration.

Experiments on PAR phenomena, performed at the REKO-3 facility at Forschungszentrum Juelich (Germany), were simulated with the CFD code ANSYS CFX. After a two-dimensional input model that represents a PAR vertical cross-section was developed, simulations were performed using experimental initial and boundary conditions. The calculated temperature axial profile along a recombiner plate was first compared to experimental measurements. After that, the temperature, velocity and hydrogen concentration fields were observed to analyze the entire process of hydrogen recombination in more detail. Finally, simulations with different initial hydrogen concentrations were performed to gain more information about possible PAR behavior at different accident conditions.

07.09.2021 10:10 Poster session RED

Severe accidents - 416

Analytical Modelling of ATF Chromium-Coated Zr-Based Cladding High Temperature Oxidation in Steam and Steam-Air Atmosphere

Alexander D. Vasiliev

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

vasil@ibrae.ac.ru

Currently, the comprehension among the specialists and functionaries throughout the world is getting stronger that the nuclear industry can encounter serious difficulties in progress in the case of insufficiently decisive measures to enhance the safety level of nuclear objects and to ensure clean energy and green world. The keen competition with renewable energy sources like wind, solar or geothermal energy takes place presently and is expected to continue in future decades. One of main measures of nuclear safety enhancement could be a drastic renovation of materials used in nuclear industry.
Up to now, several perspective advanced tolerant fuel (ATF) cladding candidates are chosen for possible application in commercial nuclear power plants (NPPs) including zirconium-based cladding with protective chromium coating, which represents more evolutional way in nuclear energy progress.
The analytical model of high-temperature oxidation of Zr/Cr cladding is developed based on oxygen diffusion consideration in the cladding. The model necessarily takes into account the initial oxidation of chromium layer with formation of chromium oxide, and, after the loss of its protective properties, the model considers the zirconium oxidation in two- or three-layers configuration. The features of Zr/Cr cladding oxidation in the steam-air atmosphere are also discussed. The model is implemented to severe accident computer running code.
The comparison of calculated results for Zr/Cr cladding high temperature oxidation with available experimental data is conducted. The reasonable agreement between calculated and experimental data is observed.
The analysis of results obtained allows make the conclusion that the application of chromium-coated Zr-based cladding may be optimistic for considerable upgrade of safety level for NPPs especially for design-basis-accident conditions.

07.09.2021 10:10 Poster session RED

Severe accidents - 419

Evaluation of containment source term of various VVER-1000/V-320 loss of coolant accidents

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

Evaluation of containment source term is an essential part of the safety analyses computational chain. This paper deals with containment fission product release during LB LOCA (DBA and DEC A) and SB LOCA calculated with COCOSYS computational code. The paper contains a description of the VVER-1000/V320 containment nodalization for thermal hydraulic and fission product transport calculation as well as a list of the most important initial and boundary conditions. The fission product release from the primary circuit includes both the fission products from fuel and from coolant and follows the principles given by the czech methodology on calculation adn determination of fission product release. The evaluation of both accidents, including the thermal hydraulic and fission product transport calculation will be presented. The main scope of this paper is the fission product release through containment leaks as well as through ventilation systems, where both filtered and unfiltered way is considered. The ratio between all investigated ways and the possible impact on the final source term and consequent radiological analyses will be discussed closely.

07.09.2021 10:10 Poster session RED

Severe accidents - 421

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

2Pro-Science GmbH, Parkstraße 9, 76275 Ettlingen, Germany

kuznetsov@kit.edu

During 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 reactor building. In presence of an ignition source, different flame propagation regimes for such a mixture may occur. The severity and a danger of such combustion process in terms of maximum combustion pressure and temperature will depend on geometry of the containment, scale, and composition of 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 liter 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.

07.09.2021 10:10 Poster session RED

Nuclear power plant operation - 503

Analysis of the effect of Krško NPP ex-core detector position on their response

Tanja Goričanec1, Marjan Kromar2, Andrej Kavčič3, Bor Kos2, Rok Bizjak3, Igor Lengar2, Božidar Krajnc3, Luka Snoj2

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

3Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

tanja.goricanec@ijs.si

During an earthquake on the 29th of December 2020 the Krško NPP automatically shut down due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation (PRNI). From the time course of the detector signal it can be concluded that there is a possibility that the fluctuation in the detector signal was caused by the mechanical movement of the ex-core neutron detectors or pressure vessel components, and not by the actual change in the reactor power. The aim of the analysis was to assess the difference in neutron flux at the ex-core detector position if the detector is moved for 5 cm in the radial or axial direction. In addition, the effect of core barrel movement for 5 mm in the radial direction was analysed. The analysis is supplemented with the thermal and total neutron flux gradient calculation in radial, axial and azimuthal direction. Monte Carlo particle transport code MCNP was used to study changes in the ex-core detector response for the above mentioned scenarios.
Power and intermediate range detectors were analysed separately, since they are constructed differently and exhibit different response characteristics. It was found that the power range ex-core detector movement has a negligible effect on the value of thermal neutron flux at the active part of the detector. However, the 5 cm radial movement of the intermediate range detector leads to ~7.3 % - 8.6 % change in thermal neutron flux within the active intermediate-range detector region. The analysis continued with the evaluation of the effect of core barrel movement on the ex-core detector response. It was determined that the 5 mm core barrel radial oscillation can lead to 8 % - 9 % change in thermal neutron flux within the active detector region. Analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, the core barrel oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

07.09.2021 10:10 Poster session RED

Nuclear power plant operation - 506

Magnetic Sludge Removal device

Jang Jinhee

New Nuclear Business Department, KHNP, 1655, Bulguk-ro Street, Munmudaewang-myeon, Gyeongju-si, Gyeongsangbuk-do, 38120 , South Korea

jangjinhee@khnp.co.kr

Magnetic sludge generated by erosion of the system piping and equipment are deposited in specific locations of major facilities of nuclear power plants such as pipes, valves, steam generators, moisture separators, heat exchangers, etc., causing corrosion of facilities or becomes the main cause of shortening their lifespan.
The need for devices to address such problems had been raised. Therefore, a Magnetic Sludge Removal device (MSRD) was developed to prevent erosion of equipment or pipes through effective removal of magnetic substances during the NPP operation.
The MSRD was designed in a vertical screen type to forms a magnetic field around the device to remove ferrite particles contained in the condensate water without interfering with the flow of systematic water during the NPP operation. The MSRD was applied to be installed at Kori unit 1&2 in Korea. One year after installing this product, it was confirmed during the overhaul of Kori unit 1&2 that magnetic sludge had been significantly removed. The amount of magnetic sludge inside the steam generator of Kori Unit 2 decreased by 21%, and the amount of magnetic impurity inside the Feedwater System of Kori Unit 2 decreased by 36%.
Lately, the MSRD was designed for installation in the condenser of Krsko Nuclear Power Plant and the impact assessment of MSRD installation on the NPSH was conducted with CFD analysis, which was verified through comparing the CFD analysis results with the experimental data of MSRD mock-up test. Then, the device was installed during the outage period in April 2021.
The device will contribute to prevent corrosion of facilities and increase the operational efficiency and cut maintenance cost by removing magnetic sludge in the feedwater system and decreasing the amount of magnetic sludge inside the steam generator.

07.09.2021 10:10 Poster session RED

Thermal-hydraulics & CFD - 608

Large interface tracking algorithm for air-water slug in turbulent flow

Jan Kren, Blaž Mikuž

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

jan.kren@ijs.si

Two-phase flows play an important role in many applications of nuclear power systems. In the two-phase flow experiments large amount of data is usually generated that needs to be classified. One of the first steps is to differentiate between liquid and gas phase in order to apply post-processing algorithms.
We are particularly interested in the application of large interface tracking algorithms to Particle Image Velocimetry (PIV) measurements, which have been obtained with high-speed camera in vertical pipe. In such geometry, the reflected light from air-water interface is deformed by refraction through different materials, which results in a low signal-to-noise ratio. Thus, dynamic mask that distinguishes between air and liquid water phase with tracer particles has great importance for the quality of the PIV measurements.
In this paper, we developed an algorithm, which is able to track the large interface of the gas phase in turbulent background flow. Firstly, the algorithm is developed on synthetically generated images of bubbles. Algorithm is further tested on data of a Taylor bubble from high-fidelity numerical simulations. Finally, the algorithm is applied on measured images of an air-water slug in turbulent background flow. Space and time complexity and accuracy of the algorithm with different settings are also presented.

07.09.2021 10:10 Poster session RED

Thermal-hydraulics & CFD - 611

AN ANALYSIS OF THE INITIAL DYNAMIC OF LARGE LOCA ACCIDENT FOR THE PERSPECTIVE DIRECT FLOW SCWR

Dmitriy Fedorov1, Vladislav Filonov2, Olexandr Kovalenko2, Yiulia Filonova2

1Engineering Technologies and Developments, 8A, Alekseya Terehina str., 04080 Kiev, Ukraine

2Joint Stock Company "State Scientific Centre of the Russian Federation – Institute for Physics and Power Engineering named after A.I. Leypunsky", Bondarenko sq 1, 249033 Obninsk, Russian Federation

flowuiz@gmail.com

This work deals with the problem of predicting the three-dimensional initial dynamics of the decompression wave propagation during a large break loss of coolant accident at promising supercritical water reactor. The foreseeing of such processes is that there are difficulties in CFD methods applying for media that changes their state from the supercritical to the two-phase during decompression. At the same time the flashing conditions occur, that are very difficult to prognosticate with mechanically equilibrium models of two-phase media, which are typical for commercial CFD packages. In addition, there are problems to form boundary conditions of the circulation pipeline rupture location.
The paper represents the results of parametric analysis of the decompression process for different initial states that can characterize the normal operation modes of the upcoming SCWRs. The evaluation of the initial dynamics of the decompression process was carried out with the specially developed representative model of a prospective direct flow reactor, which was obtained from the previously proven ANSYS CFX model of the VVER-1000, and the RELAP5 equivalent one.
The obtained outcome allows to make the conclusions about the possibility of using the existing codes. It also indicates the directions for their adaptation to analyze the processes at the supercritical state of the coolant.

07.09.2021 10:10 Poster session RED

Thermal-hydraulics & CFD - 614

Influence of seeding particles on Particle Image Velocimetry measurements in single-phase turbulent pipe flow

Blaž Mikuž1, Jan Kren1, Danjela Kuščer2, Anil Kumar Basavaraj1

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

2Jožef Stefan Institute, Electronic Ceramics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

blaz.mikuz@ijs.si

Particle Image Velocimetry (PIV) is a widely used technique for flow measurement in fluids. It is an optical method for flow visualization, which is non-invasive and, to a large degree, non-intrusive. In order to enhance the signal from reflected laser light, seeding particles are added to the fluid, which have similar density as the fluid and are of similar or smaller sizes as the smallest eddies in the flow. Ideally, such seeding particles follow the flow motion at all scales, however, in realistic flow they might affect the velocity measurement, in particular the flow statistics of higher orders.
In the present study, PIV measurements are performed in turbulent pipe flow using three types of seeding particles made of different materials, sizes and concentrations. Seeding particles made of borosilicate glass in the form of hollow spheres have diameters of 9-13 µm and density of 1.1 g/cm3. Fluorescent seeding particles are made of two different materials filled with Rhodamine B resulting in two size groups. The smaller ones are about 1-20 µm in size and made of polystyrene (PS) with density of 1.05 g/cm3. The larger ones have diameters with 20-50 µm and are made of polymethyl methacrylate (PMMA) with density of 1.19 g/cm3. Seeding particles have been added to demineralised water using various concentrations. For each sample, fluid properties have been measured. Obtained PIV measurements of mean velocity and velocity fluctuations are compared with the results of Direct Numerical Simulation (DNS).

07.09.2021 10:10 Poster session RED

Thermal-hydraulics & CFD - 617

Numerical Study on Convective Heat Transfer of Liquid Metals in Refractory High Entropy Alloy-Based Miniature Heat Sinks

Mahyar Pourghasemi, Nima Fathi

University of New Mexico Department of Mechanical Engineering, MSC 01 1150, Albuquerque, USA-New Mexico

nfathi@unm.edu

A typical straight miniature heat sink consists of several parallel microchannels fabricated on top of a base solid block. Heat sinks are usually in direct contact with hot surfaces to remove heat from them. Therefore, dissipated heat first diffuses within the base solid block and eventually is transferred to the coolant flowing within microchannels of the heat sink. This is a conjugate heat transfer phenomenon with temperature and heat flux continuity boundary conditions at the solid-fluid interfaces at the microchannel walls. Local wall temperature and heat flux distributions along microchannel walls in the heat sinks determine the local heat transfer rates and Nusselt number values. Flow and heat transfer of conventional coolants such as water and air have been extensively studied in the literature so far. Liquid metals with high thermal conductivity and high boiling temperatures are interesting coolants for applications involving high working temperature ranges. Low Prandtl numbers of liquid metals result in relatively small Peclet numbers for their flow within miniature heat sinks. Heat conduction can play a significant role in low Peclet number flows in miniature heat sink by altering local wall temperature and heat flux distributions. Moreover, liquid metals such as Na are corrosive and react with most commonly used high thermal conductivity solid materials such as copper. To address this problem, we are modeling flow and heat transfer of Na and NaK in Refractory High Entropy Alloys (RHEAs)rectangular minichannel heat sinks. The results are compared against Inconel-based microchannel. The Effect of solid base material thermal conductivity on the heat flux and temperature distributions over the heat sinks walls is investigated. Local and average Nusselt numbers are calculated and compared with each other for heat sinks with different base solid materials of RHEAs.

07.09.2021 10:10 Poster session RED

Thermal-hydraulics & CFD - 620

Validation of two-fluid boiling flow model on the DEBORA benchmark experiments

Boštjan Končar, Matej Tekavčič, Andrej Prošek

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

bostjan.koncar@ijs.si

With continuous development of three-dimensional multi-phase modelling capabilities, the numerical simulation of complex two-phase flows becomes feasible. In particular, the two-fluid model relying on phase-averaged equations has gained a lot of attention in the past decade due to its wide range of applicability and above all, as the model applicable to industrial conditions. However, the two-fluid model approach requires many closure relations and a great diversity of sub-models has been developed. So far, it seems that none of them has reached a general level of applicability to any flow condition. It is clear that each set of sub-models needs to be validated against small-scale experimental data. In most cases, the authors use its own validation database, therefore the range of validity of their models is difficult to compare. An effort in this direction has been made in the frame of NEPTUNE project, supported by Commissariat a l'Énergie Atomique et aux Énergies Alternatives (CEA), Electricité de France (EDF), Framatome and Institut de Radioprotection et de Sureté Nucléaire (IRSN), where the benchmark test has been launched trying to pave the way towards a unified method for testing and validation of two-fluid closure models based on publicly available experimental data. The first tests will be focused on high-pressure flow boiling in a simple tube geometry, performed in DEBORA experimental facility at CEA-Grenoble. The DEBORA experiments provide a reliable database on local measurements of boiling phenomena in a simple vertical tube geometry with electrically heated wall. A turbulent boiling flow of Freon R12 or R134a has been used to mimic the high-pressure conditions, relevant for nuclear applications in Pressurized Water Reactors (PWR). The DEBORA benchmark will provide to the Computational Multiphase Fluid Dynamics (CMFD) community a large database of available experimental data and prepare the validation of simulation results of participants in an organised way. In the second phase, new data will be provided and used for blind tests.
The Reactor engineering division of Jožef Stefan Institute decided to participate in this benchmark, taking advantage of the many years of experience with modelling of different types of multi-phase flows. In this work, the predictions of available DEBORA boiling tests will be presented. The prediction capability of two-fluid boiling flow models in the Ansys Fluent code will be examined. Boiling on the heated wall will be modelled by the heat-partitioning model that includes several sub-models for bubble departure diameter, nucleation site density, bubble detachment frequency etc. Besides, interfacial coupling effects between the vapour and liquid phase and turbulence effects will have to be included. The simulation results will be compared with the local measurements of radial profiles of void fraction, interfacial area density, bubble size and liquid temperature.

07.09.2021 10:10 Poster session RED

Nuclear regulations - 705

25 Years of the Slovenian Participation in IAEA ITDB Programme and Endeavours to Combat Illicit Trafficking of Nuclear and Radioactive Material

Janez Češarek

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

janez.cesarek@gov.si

Back in 1995, the International Atomic Energy Agency (IAEA) launched its efforts to systematically collect and analyse information from its Member States, addressing the so-called illicit trafficking of nuclear and radioactive material. Nuclear trafficking was clearly in the limelight due to several, well published and echoed seizures of enriched uranium and plutonium in Europe at that time. Collected information in the Illicit Trafficking Database (ITDB) has become one of the valuable “flagships” of IAEA in the sphere of nuclear security. The name of the database was changed in 2013 to be: Incident and Trafficking Database.

Slovenia joined the ITDB in October 1995, thus marking last year the 25th anniversary of its participation in the database. The membership in this voluntary database has been growing steadily since its inception. The Slovenian Nuclear Safety Administration (SNSA) is the national point of contact – also taking care of national reporting into ITDB as well as distributing information from it – based on a need-to-know basis. Years ago, SNSA established an informal group on combatting illicit trafficking of nuclear and radioactive material – being a kind of a vehicle to deliver and exchange pieces of pertinent information.

So far, and fortunately, there has not been detected (or intercepted) any deliberate trafficking of nuclear (radioactive) material in Slovenia which was the case in Europe particularly in the “turbulent 90ies”. However, there is no room for complacency and past years have nevertheless brought up a number of reported cases – mostly the so-called “orphan sources”, spanning from U- and Th-substances to Kr-85, Cs-137 and other sources, including depleted uranium. The article will wrap up the current Slovenian criteria for reporting into ITDB. Moreover, it will touch upon the role of ITDB-related data as a source of information into the national (nuclear) threat assessment “process”. In addition, some extra thoughts will be presented on other databases and open-source information and how to “amalgamate” a broader picture, trends, patterns and conclusions. There is a will to nurture a pro-active approach, learn from the past cases – and the ITDB-related data as well as other open-source information are important remits for tailor-made national outreach activities.

Slovenian counterparts have been active and vigilant also in those issues which have not been related to trafficking and malicious intention but rather to inadvertent movement of e.g. radioactive sources in scrap metal (including contaminated items in semi- and final products) which may “travel” through more than just one continent – as some less or more known cases showed in the past years. The recently issued Decree on checking the radioactivity of consignments that could contain orphan sources (being fully mandatory since March 2020) has been enshrined as a fulcrum for the improvement of detection capabilities at the major Slovenian nodal points.

International outreach has been multi-pronged, e.g. through the IAEA/ICTP Nuclear Security School in Trieste (Koper), participation in different regional (European) engagements and topical meetings at the IAEA headquarter. There have been adjacent efforts, e.g. by US-counterparts to enhance detection and response in this sphere. Awareness raising was an underlying issue to spur up further pro-activeness. It is noteworthy to add that in 2018, Slovenia joint two international initiatives in the area of nuclear security; one being INFCIRC/918 (“Joint Statement on Countering Nuclear Smuggling”). To conclude, SNSA as the Slovenian nuclear regulator has pursued a number of activities to prevent, detect and response to illicit trafficking or curb other unauthorised activities.

07.09.2021 10:10 Poster session RED

Nuclear fusion and plasma technology - 804

Neutronics analyses of EU DEMO 2020 EC port configuration

Aljaž Čufar1, Christian Bachmann2, Thomas Berry3, René Chavan4, Tim Eade3, Thomas Franke5, Bor Kos6, Dieter Leichtle7, Peter Spaeh7, Tran Minh Quang4

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

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

3Culham Centre for Fusion Energy, Abingdon, Oxon, OX14 3DB, United Kingdom

4Swiss Federal Institute of Technology (EPFL), Station 3, Lausanne, Switzerland

5Westinghouse Electric Belgium S.A., Rue de I´Industrie, 43, 1400 Nivelles, Belgium

6Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

7Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany

aljaz.cufar@ijs.si

The DEMOnstrational fusion power plant (DEMO) is being developed within EUROfusion and one of the challenges is the integration of all the systems into a tokamak fusion reactor by designs that meet strict design criteria required for safe and reliable long-term reactor operation. Neutronics analyses are required as an important contribution to this effort.
Openings foreseen for the launchers of electron cyclotron (EC) heating in equatorial port plugs needed for plasma heating and control represent a challenge in terms of neutron shielding. Both the neutron streaming through the EC port and nuclear loads in its critical components have to be considered in the system design and integration.
The latest pre-conceptual EC port design was analysed in terms of nuclear loads in exposed parts of the system as well as the effect of such EC port plug penetrations on the nuclear heating of the superconducting toroidal field coils. The current design is a result of design evolution influenced by both neutronics and other considerations, e.g. structural requirements, maintainability, and plasma heating and control. The design is based on a combination of an optical mirrors system in the EC port plug with fixed and steerable mirrors, powered by gyrotrons in order to perform functions of plasma heating and control. This new design for DEMO is also called mid-steering launcher design. It was chosen as a compromise between the previously analysed remote-steering design and the ITER-like front steering design. The former was problematic in terms of neutron streaming through waveguides and a very complex integration while the nuclear loads in steering mirrors of the latter concept would lead to system’s unacceptably short lifetime.
In this paper we describe the neutronic studies for the DEMO EC mid-steering equatorial port design, i.e. nuclear heating and neutron damage assessments in important parts of the system, nuclear heating in superconducting toroidal field coils, and shutdown dose rates in accessible port areas relevant for maintenance and inspection.

07.09.2021 10:10 Poster session RED

Nuclear fusion and plasma technology - 807

Kinetic-Fluid Coupling Time-Dependent Simulations Of ITER During ELMs

Ivona Vasileska1, Xavier Bonnin2, Leon Kos1

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

2ITER Organization, Cadarache Centre, 13108 St. Paul lez Durance, France

Edge Localized Modes (ELMs)-induced transient heat loads on the divertor targets represent a important threat to target lifetime and can lead to the need to replace them with a frequency that has a major impact in the execution of the ITER Research Plan. Predicting the impact of such large transient heat loads through modelling is especially challenging and is often attempted through the use of fluid plasma boundary modelling codes, such as SOLPS-ITER, in which the ELM is crudely approximated as a fixed large, but limited in time, increase in anomalous cross-field transport coefficients for particles and heat to mimic a specified total ELM energy loss. However, one problem with this approach is that the boundary conditions at the target sheath interface are expected to vary strongly in time through the ELM transient, whilst fixed kinetic heat flux limiters are typically applied in the fluid codes. Coupling kinetic fluid codes has not yet been systematically used for ITER ELMs study.

This contribution describes the first results of efforts to address ELMs issues for ITER simulations under high performance conditions using the 1D3V electrostatic parallel Particle-in-Cell (PIC) code BIT1, to study the kinetic effects and to provide time dependent kinetic target sheath heat transmission factors (SHTF). In a later stage of the work, these will be used in the formulation of fluid boundary conditions for calculations of ELM target heat loads using the SOLPS-ITER code.

The BIT1-SOLPS-ITER coupling allows us to investigate the kinetic effects on the targets, by comparing power and particle fluxes from time-dependent simulations of ITER Type I ELMs.

07.09.2021 10:10 Poster session RED

Nuclear fusion and plasma technology - 810

Effects of the source properties on the filamentary transport in the tokamak SOL

Jernej Kovačič1, Stefan Costea2, Tomaz Gyergyek1

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

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

jernej.kovacic@ijs.si

We are expecting first plasma in the largest experimental fusion reactor ITER in a few years and we are already in the design phase of the future demonstration power plant (DEMO) reactor. These two machines will behave vastly different than the current tokamaks because of their size and the amount of the energy that will be stored in the plasma. The plasma heat and particle loads to the first wall and the divertor will be crucial in determining the lifetime of these tokamaks as well as their performance. The currently available materials for the plasma-facing-components (PFCs) will be a major factor, as they will only be able to withstand the limited heat and particle loads which cause ageing and erosion. Since drastic improvements of materials cannot be expected, the solution is to have particle and energy fluxes well regulated and for that they have to be thoroughly understood.
There are several possible scenarios for plasma transport in the SOL of future tokamaks and all of them are avoiding the possibility of the destructive type-I ELMs (edge-localized-modes), which are a standard of the today’s most common H-mode of operation. The major relaxations that these pedestal crashes provide are replaced by smaller events, namely by various types of filaments securing most of the transport across the SOL.
In our work we build on our previous model of blob-filaments by a fully-kinetic particle-in-cell (PIC) code BIT1. The model involves fully-kinetic charged and neutral particles, with arbitrary sources of filaments that constitute the SOL. The particle and energy sinks are self-consistently calculated in the parallel direction, while a diffusion module accounts for the radial losses. Neutrals are introduced either as a consequence of divertor and first wall recycling or through a volumetric source mimicking gas fuelling/puffing. There are most of the relevant collisional processes included in the simulations such as ionization, excitation and elastic collisions, as well as plasma-wall-interactions, both of which are operated through a rich Monte-Carlo collisional module with exact cross-sections.
The present work focuses on the role of the particle and energy source on the transport properties. There are several modes of SOL operation, such as the I-mode the quasi-continuous exhaust (QCE) and the enhanced D-alpha (EDA), with inter-ELM standard H-mode also still being explored. We studied various cases by a parametric study of the following quantities: the filament amplitude, duration (FWHM – full-width at half-maximum) and the waiting time between the filaments as well as the energy distribution functions of the injected particles. The source properties show to have a significant impact on the deposition of energy in the divertor, the changes of the ratios of parallel to radial transport and the position of the ionization front.

07.09.2021 10:10 Poster session RED

Nuclear fusion and plasma technology - 813

Thermionic emission from the divertor surface during energetic events in tokamaks

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

jernej.kovacic@ijs.si

Thermionic emission is a common phenomena in plasma technology, as it is often used as a source of free electrons. It is described by the Richardson-Dushman law for general conditions, and by Child-Langmuir law in the space-charge-limited regime. Thermionic emission is, however not a desired process in fusion plasmas, since it is achieved only close to the melting temperatures of the materials. Until now such emission is only observed on a small scale, e.g. with electrical probes being plunged too deep into the confined plasma [1]. With future tokamaks being significantly larger and having higher densities and temperatures in the edge plasma, events such as thermionically emissive divertor have a much higher probability. Last year we have started developing a PIC model of the scrape-off-layer plasma coupled to a thermal model of the boundary material. This model has now been enhanced and we have been able to study the phenomena in much more detail.
We have introduced a new time dependent thermal model with temperature dependent material properties based on fusion material database [2]. Our fully-kinetic approach allowed us to also model a very detailed source of particles and energy corresponding to type-I ELMs and other energetic events. The thermionic emission presents itself as a significant cooling mechanism for the material surface, as each emitted particles takes away at least the energy equivalent to the work function of the material. This effect could be seriously reduced, if we introduced an oblique magnetic field into the model, as prompt redeposition of emitted electrons was substantial. We have also learned that the EM radiation of the surface due to its high temperature plays an important role in the cooling process even before the emission starts and often prevents it.
In our simulations we have made a parametric scan of several quantities including, but not limited to, plasma source parameters (density, temperature, event duration) and external magnetic field.
Our diagnostics involve temporal evolutions of plasma profiles of macroscopic quantities, plasma particle and energy fluxes and surface temperature.

[1] J. Adamek et al., Contrib. Plasma Phys. 54 (2014), pp. 279-284
[2] P. Tolias, Nucl. Mat. Energy 13 (2017), pp. 42-57

07.09.2021 10:10 Poster session RED

Materials and ageing management - 905

The Regulatory Oversight of the Reactor Pressure Vessel Suitability for the Krško NPP Long Term Operation

Tom Bajcar, Andreja Peršič, Sebastjan Šavli

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

andreja.persic@gov.si

The reactor pressure vessel (RPV) is one of the most important nuclear power plant passive components. Based on the reactor pressure vessel embrittlement, the operating regime is determined. During the plant operation, particular attention shall be paid to the RPV maintenance, testing and in-service inspection. The specific programs and related implementation procedures are in place to do these. Monitoring of the operational experience, exchange of information in the context of international initiatives, and international independent reviews further contribute to the effective supervision of the RPV's condition.
Normally, the RPV's design, material composition and the operating conditions determine its quality and suitability for the extended life time operation of the nuclear power plant. To extend the operational life time of the Krško NPP for additional 20 years, a number of safety improvements of SSCs, including RPV have been made. The replacement of the reactor pressure vessel head (based on the experience from the Davis Besse NPP event) is one of the most significant examples. The ageing management program was prepared for the NPP life time extension. This was based on the new ex-vessel dosimetry program that was one of the prerequisites for the Krško NPP ageing management program approval, as well as on the results of periodic in-service inspections and strict monitoring of operational transients which can cause fatigue of RPV’s elements.
International activities related to the RPV, i.e. the basic material manufacturing issue, ENSREG international review-Topical Peer Review on the Ageing Management, are just some of the activities in the last decade where the Slovenian Nuclear Safety Administration has also played an active role.
The RPV issues related with the Krško NPP extended life time operation will be discussed in the article with the emphasis on the SNSA regulatory activities.

07.09.2021 10:10 Poster session RED

Materials and ageing management - 908

Review of M5™ Cladding Models Relevant for LOCA Simulation with the TRANSURANUS Code

Rolando Calabrese1, Arndt Schubert2, Paul Van Uffelen2

1ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy

2European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Hermoltz-Platz 1, 76344 Eggenstein-Leopolshafen, Germany

rolando.calabrese@enea.it

Reliable and accurate fuel performance codes are crucial for the design of nuclear fuel. For this purpose codes should have the capability of predicting fuel behaviour under both normal operation and postulated accidents. Phenomena occurring under these two conditions are quite different and this fact has led to the development of distinct codes for normal and accident conditions. With this regard, the TRANSURANUS code is featured by a clearly defined mechanical and mathematical framework which has permitted since its beginning a continuous development and an extension of the domains of its application.
In particular, strong efforts have been devoted to make the code applicable to LOCA transients in LWR rods. Besides newly developed correlations for off-normal conditions and an increasing detail in the definition of boundary conditions, new cladding materials have been considered. Moving from standard Zry-2 and Zry-4, the material correlations for E110 and more recently for M5 alloys have been introduced in the code. Within the Reduction of Radiological Consequences of design basis and design extension Accidents project (R2CA) of EURATOM, ENEA, in coordination with the JRC, will consider the topic of hydrogen uptake for M5, conducting in parallel a review of M5 models relevant for LOCA simulations such as phase transition and high temperature creep. Results are presented in this paper.

07.09.2021 10:10 Poster session RED

Environment and back end of the fuel cycle - 1004

Neutron Shielding Performance Analysis of the Ordinary Concrete Reinforced with an Innovative H3BO3 Polymer Based Particles

1Islamic Azad University, Department of Nuclear Engineering, Science and Research Branch, Hesarek, 1477893855 Tehran, Iran

2Department of Nuclear Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran, 1477893855, Iran

In the present study, an innovative Boron based polymer composite was proposed and synthesized. This polymer is based on the H3BO3 as the main neutron absorber which is sandwiched between polymer layers. The characterization results showed that the particles are sphere and strong bands are available between H3BO3 and hydrocarbon elements. The Monte Carlo simulation of the ordinary concrete reinforced with different size and weight fraction of the polymer composite has been performed. The results showed that the nano particles are more effective in comparison with micro ones and the neutron transmission ratio was reduced by decreasing the size of the H3BO3 polymer particles. The self-shielding effects as well as the collision probability between thermal neutrons and H3BO3 polymer particles are two important phenomena for this behavior. The neutron efficiency of prepared samples was estimated for thermal (0.025eV), epithermal (0.2 eV), intermediate (1keV) and fast (10 MeV) neutrons. The neutron absorption cross section was increased by increasing the filler content and decreasing the neutron energy. It was demonstrated that the use of H3BO3 based polymer nano particles as filler in ordinary concrete shielding materials would enhance the radiation property of the shield.

07.09.2021 10:10 Poster session RED

Environment and back end of the fuel cycle - 1008

INVESTIGATION OF OCCUPATIONAL RADIATION EXPOSURE FROM C-ARM FLUOROSCOPY GUIDED PROCEDURES

Turkki Shatona

Ministry of Health and Social Services - National Radiation Protection Authority (NRPA), Harvey Street, Ministry of Health and Social Services, ministerial Building, 9000, Windhoek, Namibia

turkki.shatona@mhss.gov.na

The use of fluoroscopy as a medical imaging technique in the diagnoses and treatment of many diseases is one of the main source of occupational radiation exposure. Scattered dose rate around a fluoroscopy room, to determine staff exposure during C-arm fluoroscopy guided procedure to measured and analysed. The scattered radiation dose was measured at 12 positions across the operating theatre using a radiation survey meter. A Perspex phantom to simulate a patient was placed on a C-arm x-ray machine. Staff radiation dose measurements were conducted in the same operating theatre where three staff (doctor, nurse and radiographer) were monitored using thermoluminescent dosimeters (TLDs) over a period of three months. Both staffs wore TLDs at the neck, chest and waist over the lead apron.
The scattered radiation dose decrease with distance from the source. This is very important on reducing scattered radiation to staff during fluoroscopic procedures along with the use of proper shielding and less exposure time. The dose received by staff over a period of three months is less among the nurse and radiographer while the doctor received the highest. Reason being that the doctor is always in close proximity to the patient during C-arm fluoroscopy guided procedures.

07.09.2021 10:10 Poster session RED

Environment and back end of the fuel cycle - 1011

Adsorption of Iodine on Painted Surfaces in Nuclear Power Plants Containment Buildings

Iana Zamakhaeva

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

iana.zamakhaeva@ujv.cz

The paper is focused on problematics of adsorption and desorption of iodine on painted surfaces in containment buildings during loss-of-coolant accidents. The theoretical part of the paper contains an introduction into iodine behaviour in containment buildings, short description of the COCOSYS (COntainment COde SYStem) computational code used for analysis and a description of validation chain defined by the IAEA. The experimental part of the paper is aimed at validation of COCOSYS iodine paint deposition model in dry conditions and fine-tuning parameters of this model on selected experiments executed under the OECD NEA Behaviour of Iodine Project. The model fine-tuning aims at adsorption and desorption rate as well as the splitting factor of physisorbed and chemisorbed iodine. The selected experiments include those with Ameron Amerlock paint, which is present in the Temelin WWER-1000/V320 NPP containment. This work was conducted in the framework of the EU R2CA project.

07.09.2021 10:10 Poster session RED

Environment and back end of the fuel cycle - 1014

Neutronic Analysis of Ventilated Dry Storage Cask with Monte Carlo Method

Burak Sarioglu

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

buraksarioglu@itu.edu.tr

Turkey is constructing its first nuclear power plant of VVER-1200 type. The necessity of beginning of planning on nuclear waste management at the early stages of the construction is a known fact. In the meantime, dry storage casks are widely used worldwide for interim storage until final repository is on-line. Therefore, it is very important to build capacity on neutronic, thermal-hydraulic, and mechanical analysis of dry storage tanks. This study focuses on neutronic analysis of air ventilated dry storage cask design VSC-24 which is in use for VVER-1000 type reactor spent fuels. The Monte Carlo method was selected as simulation tool. Spent fuels with different burnup and pool residence times were considered. In addition, different options of filler gases in the steel basket were examined as well as different basket thicknesses. Depending on the scenario at hand, the spent fuel composition therefore the source term was calculated for Monte Carlo simulations. Dose rates consisting of gamma and neutron radiation were calculated at different locations on the cask. The simulation results are in good agreement with the data at the literature. Therefore, the capacity of modeling and simulating dry storage tanks is enhanced.

07.09.2021 10:10 Poster session RED

Environment and back end of the fuel cycle - 1017

The impact of neutron irradiation on concrete structures

Szabina Török1, Sugár Viktória2, Lama Alnatour3

1Centre for Energy Research Hungarian Academy of Sciences, P.O.Box 49, H-1525 Budapest, Hungary

2Hungarian Academy of Sciences Centre for Energy Research, Budapest 114, P.O. Box 49, Hungary, H-1525, Hungary

3Óbuda University, Ybl Miklós Faculty of Architecture and Civil Engineering , Thököly út 74, 1146 Budapest, Hungary

torok.szabina@energia.mta.hu

Concrete is the most common and the predominant material used in the construction of nuclear plants, radioactive waste repositories and their shielding materials. Many of these structures are large and irreplaceable sections; therefore, studying and enhancing the durability of the concrete design overtime is essential. The microstructure and properties of dry cement paste and/or aggregates of different types change over time due to slow hydration, crystallization of amorphous constituents, and reactions between cement paste and aggregates as well as influences from the local environment such as humidity, temperature, radiation exposure, and/or chemical interactions. This paper is reflecting the current state of knowledge about the components of concrete and the behavior under such circumstances over time.

07.09.2021 10:10 Poster session RED

Environment and back end of the fuel cycle - 1020

Decommissioning of activated fragments of the Jaslovské Bohunice V1 NPP VVER-440 pressure vessel

Stefan Cerba, Milan Myslík, Branislav Vrban, 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

stefan.cerba@stuba.sk

With the construction of new nuclear power sources, the issue of decommissioning of nuclear power plants has also come to the fore for a considerable time. During the decommissioning of a nuclear power plant, decisions need to be made that are often influenced by the specific decommissioning facility. The decommissioning of the V1 NPP in Slovak Republic has arrived into the phase of fragmentation of the reactor pressure vessel and storage of resulting fragments. For this purpose, it is always necessary to perform analyses dealing with storage options. This paper is dealing with the possibilities of the decommissioning of these components. The first part is focused on the categorization of reactor pressure vessel fragments, based on various criteria. In the next part, analyses aiming on the definition of the source term and the possibilities of fragments storage are presented. After determining the initial conditions, the main part of the work deals with the estimation of dose rates in the vicinity of fiber-reinforced concrete containers and storage conditions in the Integral RAW Storage Facility using the SCALE6 code system. As the most important result of the work, the number and type of fiber-reinforced concrete containers are presented, which are necessary for storing the entire reactor pressure vessel.

07.09.2021 10:10 Poster session RED

Education, training and outreach - 1105

Public Opinion about Nuclear Energy – Year 2021 Poll

Retired from JSI, Jamova 39, 1000 Ljubljana, Slovenia

An important activity of Nuclear Training Centre at the Jožef Stefan Institute is informing the general public about nuclear power and nuclear technology, about radioactivity, about Krško Nuclear Power Plant and about energy in general.
Our main target population are schoolchildren from the last grades of elementary school and from high school (ages 13-18) with their teachers. 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. In the last decade, we had close to 8000 visitors per year, but in the years 2020 and 2021 the number of visitors was significantly lower number due to limitations imposed by covid-19 pandemic. Also, instead of using a classroom, all the lectures and demonstrations of radioactivity experiments were given by videoconference means. Furthermore, a virtual tour of our permanent exhibition and a virtual tour of the TRIGA research reactor were offered. .
Since 1993, we monitor the opinion trends by polling some 1000 youngsters, i.e. our visitors. 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. In the last two years, the number of collected questionnaires was lower, but nevertheless meaningful results were obtained.

07.09.2021 10:10 Poster session RED

Education, training and outreach - 1108

Sandi Cimerman1, Bojan Žefran2

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

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

bojan.zefran@ijs.si

The article will describe the upgrade of the HPCC (High Performance Computing Clusters) shared computing infrastructure at the JSI Reactor Center in the last decade.
HPCC have become essential in the nuclear energy field. They are used for calculations in neutronics, thermalhydraulics, material science and structural integrity. All these calculations improves safety and optimize use of a nuclear installations.
The Reactor Engineering Department (R4) and the Reactor Physics Department (F8), with the support of The Slovenian Research Agency (ARRS), have been jointly operating their own High Infrastructure Computing Clusters for more than fourteen years. The two departments have invested heavily in HPCC systems during this time, as these represent critical equipment for their research activities. During this time, purchases and multiple updates of several separate clusters were carried out: Mangrt (2007/2008), Krn (2010/2011/2012), Razor (2014/2014 // 2015) and Skuta (2019/2020/2021). The departments already have all the key knowledge necessary for the operation of the clusters, administration, maintenance, as well as direct support to users. Over the last decade, the sections have also significantly modernized and upgraded the existing infrastructure. This allows us to safely and reliably operate existing, as well as easily install new computing capacity. The departments take special care of security and access to sensitive existing data and software packages.
The main focus in the future will be to ensure the effective scalability of parallel computing, calculations, data and programs and to support reactor calculations at both departments.

08.09.2021 10:10 Poster session GREEN

Advances in nuclear technology - 205

Development of technical decision for digital support of the NPP construction using photogrammetry methods

JSC "Proryv", Malaya Krasnoselskaya street, 2/8, 107140 Moscow, Russian Federation

The «Proryv» project implemented by the State Atomic Energy Corporation ROSATOM is aimed at achieving a new quality of nuclear energy, development, creation and industrial implementation of closed nuclear fuel cycle (CNFC) based on fast reactors that will lead to development of a large-scale nuclear power industry.
Drawings and 3D-models of nuclear facilities are created during the implementation of this project. Deviations arise during the construction of nuclear facilities and their filling with equipment. This study provides a methodology for finding deviations between design and constructed objects based on photogrammetry algorithms. The SfM-algorithm (Structure from Motion) as a computer vision algorithm is commonly applied and has been extensively studied. The possibility of using the SfM-algorithm during the construction of the NPP (nuclear power plant) is considered in this study.
The aim of the study is to analyze basic 3D-scanning methods to find a decision for digital support of the NPP construction. This study demonstrates the features of applying the SfM-algorithm to control the NPP construction.
3D-models of nuclear facilities were obtained by using SfM- and MVS-methods and then comparing these models with the design (with the reference) made it possible to find construction deviations. Important characteristics of the methodology are accuracy, usability and profitability.
The technique for digital support of the NPP construction was found and tested. The obtained technique has already found its application in the NPP construction as a part of the «Proryv» project in Russia.
The study enables a better understanding of scanning methods advantages and the scope of their application.

08.09.2021 10:10 Poster session GREEN

Advances in nuclear technology - 208

Uncertainties and Sensitivity Analyses of ULOF and UTOP in Sodium-cooled Fast Reactor

Rolando Calabrese

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

rolando.calabrese@enea.it

Next generation fast breeder reactors (FBRs) aim at improving nuclear energy sustainability by overcoming, among others, the limitation in uranium resources. Under irradiation conditions that are typical of FBRs fuel pins face a significant degradation of thermophysical properties and microstructural changes that limit the residence time in reactor. Innovative cladding materials have been proposed with the objective of extending the lifetime of fuel.
In the frame of the Working Party on Scientific Issues of the Fuel Cycle, the Expert Group on Innovative Fuel Element (WPFC/EGIFE) has organized a follow up of the first benchmark on fuel performance codes. Phase II will deal with simulations of loss of flow and over power unprotected transients in fast reactors (ULOF, UTOP). The TRANSURANUS code will be used for the analysis of an oxide reactor core cooled by sodium. The fuel pin is featured by the use of an innovative cladding material: the 9Cr-ODS alloy. The TRANSURANUS team has devoted significant efforts to improve and refine the models used for fast reactor fuel pins, therefore, the analyses proposed in this benchmark are a good opportunity to test the new capabilities of the code. A deterministic analysis of the irradiation history together with a statistical analysis of the transients is presented in this paper. This latter study will be conducted by means of the TUPython module that has been recently released to the users.

08.09.2021 10:10 Poster session GREEN

Advances in nuclear technology - 211

Neutron flux measurement in TEPLATOR DEMO

Eva Vilimova, Tomas Peltan, Radek Škoda

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

vilimova@kee.zcu.cz

The design of the new reactor core goes hand in hand with the design of an adequate neutron flux measurement to ensure power control and main safety functions. The conceptual design of the new proposed nuclear reactor for heat production, the TEPLATOR, will be supplemented by a proper in-core and ex-core neutron flux detection. The focus of this work is mainly on ex-core neutron flux measurement. This article builds on the previous research concerned with a verification of the possible use of the special neutron measuring system placed in a graphite reflector of the TEPLATOR. More calculations were carried out with a Monte Carlo code Serpent, which serves as a calculation tool for the purposes of this article. Further research deals with the optimization of the proposed radial and axial layout of the ex-core detectors in the graphite reflector of the TEPLATOR DEMO. The layout is adapted to a core power distribution, which is studied. This work is also focused on neutron flux and spectrum changes during reactor operation.

08.09.2021 10:10 Poster session GREEN

Reactor physics & research reactors - 309

Parametric analysis of the closed-water activation loop at the JSI TRIGA reactor

Domen Kotnik1, Kristina Pahor2, Luka Snoj1, Igor Lengar1

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

2Fakulteta za matematiko in fiziko, Jadranska 19, 1111 LJUBLJANA, Slovenia

domen.kotnik@ijs.si

Water as a primary coolant is being used in most of today's fission reactors and it is also one of the most promising coolants for future fusion reactors. During cooling of the reactor blanket (fusion)/core (fission), the water is exposed to neutrons, produced in fusion/fission reactions, and gets activated, which leads to radioactive decay with releasing high-energy gamma rays and neutrons. Many computational analyses of the water activation process have been performed for ITER and DEMO; however, results are subject to enormous uncertainties and consequently poor quality due to lack of experimental nuclear data, inaccurate computational methodologies/codes and lack of experimental facilities for experimental validation of methodology.
In light of this, we will design, manufacture and commission a water activation loop at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor. To do that, the current research reactor JSI TRIGA will be upgraded by constructing a closed-loop for water activation, which will serve as a well-defined and stable 6 MeV – 7 MeV gamma-ray source. The preliminary design of the irradiation facility [1] will be optimised by using the particle transport models and considering the time-dependent radiation sources. This includes all engineering necessities and the neutronics/radiation specifications (e.g., optimal position and shape of activation part, position, set-up and sensitivity of neutron and gamma detectors). In the design, special attention will be devoted to the estimation and reduction of experimental uncertainties with the aim to design a facility for performing benchmark experiments. In order to design an optimal water activation loop detailed uncertainty and sensitivity study of the most relevant parameters, e.g. irradiation time, loop time, transport time to the detectors, reaction rates and decay constants will also be performed.

[1] Žohar A., et al., Conceptual Design of Irradiation Facility with 6 MeV and 7 MeV Gamma Rays at the JSI TRIGA Mark II Research Reactor, EPJ Web of Conferences 225, 04014 (2020), https://doi.org/10.1051/epjconf/202022504014.

08.09.2021 10:10 Poster session GREEN

Reactor physics & research reactors - 312

Non-Modal Stability Analysis of the Zero-Dimensional Model of the TRIGA Mark II Reactor

Carolina Introini1, Antonio Cammi2, Parikshit Bajpai1

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

carolina.introini@mail.polimi.it

Non-modal stability theory is a new approach to stability analysis focused on the short-term behaviour of the system following a perturbation. Traditional modal techniques focus instead on the asymptotic response, potentially missing finite-time instabilities and temporary growth of parameters such as temperature and mass flow rate. Although widely used in the fluid-dynamics analysis, this technique has been only seldom applied to nuclear reactors. Building upon previous works done by the author, in the following the non-modal stability theory is used to study the stability of the zero-dimensional complete model of the TRIGA Mark II reactor, focusing on the short-term response to perturbations (power, inlet temperature). The primary aim of this work is to show how the widely-used modal stability techniques fail to predict transient growth of the perturbation for asymptotically stable systems, which, if repeated, could potentially lead to additional wear and tear of the system.

08.09.2021 10:10 Poster session GREEN

Reactor physics & research reactors - 315

The Effect of Burnup on ITU TRIGA Mark II Research Reactor Control Parameters

Fadime Ozge Ozkan1, Senem Senturk Lule2, Uner Colak2

1Istanbul Technical University, Maslak, 34467 Sariyer/Istanbul, Turkey

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

ozge.ozkan@itu.edu.tr

The fuel temperature, control rod worth, pool water temperature, reactor power, and prompt temperature coefficient are important concepts to define research reactor system design basis. In this study, change in control rod worth, excess reactivity, neutron flux, and delayed neutron fraction of ITU TRIGA Mark II Research Reactor was calculated for 55 days burnt fuel. 3D full core MCNP 6.2 model was used for burnup calculations to obtain the fuel composition in each of the 69 fuel elements. The new core configuration was modeled to create integral rod worth curves of transient, safety, and regulating rods numerically. ENDF/B-VII library was used for the cross section data and rod insertion method was employed for numerical analysis of control rod worth. The rod worth calculation methodology, MCNP model and simulations were benchmarked against the fresh core configuration results for validation and verification. The results showed the change in safety related parameters. Especially, the rod worth and excess reactivity are decreased with the burn-up. The new core configuration is necessary to receive the lost reactivity due to burnup by shuffling and fresh fuel addition.

08.09.2021 10:10 Poster session GREEN

Reactor physics & research reactors - 318

Gathering of Data on the European Research Reactor Fleet as Part of the TOURR Project

Bor Kos1, Roberta Cirillo2, Anze Pungercic1, Pavel Gabriel Lazaro2, Luka Snoj1

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

2European Nuclear Education Network, Rue d’Egmont 11, 1000 Brussels, Belgium

bor.kos@ijs.si

The primary objective of the TOURR project is to develop a strategy for Research Reactors (RRs) in Europe and prepare the ground for its implementation. This strategic goal can be divided into specific objectives: Assessment of the current status of European research reactors fleet, including plans for upgrades, evaluation of urgent EU needs, developing tools for optimal use of the research reactors fleet and finally, rising awareness among decision-makers on the (future) role of research reactors.

The ambition of the TOURR project is to secure access and availability of RRs as a vital part of the European Research Area and support a stable supply of medical radioisotopes. Nuclear RRs have been constructed in countries implementing nuclear power plants and used in experiments necessary to develop commercial reactors and research&training programmes.

The first part of the project, led by JSI, aims to collect and update information on the European RR fleet and their plans in the period 2020-2030. Furthermore, gap analyses of the RR will be performed in the areas of science & technology, medical and industrial radioisotope production and education & training.

The paper will introduce the TOURR project and the specifics regarding the first step of the project – gathering data on the European RR fleet. The paper will show the process of creating and distributing a dedicated questionnaire used to gather relevant information on the current status and plans and needs of the European RR fleet. Additionally, preliminary observations on the successful deployment of the questionnaire and lessons learned will be presented.

ACKNOWLEDGMENTS
This project receives funding from the EURATOM Research and Training programme 3 years under grant agreement N° 945269.

TOURR CONTRIBUTORS
See list of TOURR contributors in: G. L. Pavel et.al., NENE 2021.

Footnote
Bor Kos, Currently at Oak Ridge National Laboratory, Tennessee, USA.

08.09.2021 10:10 Poster session GREEN

Reactor physics & research reactors - 321

Burnup-dependent isotopic compositions of PWR fuel pins using OpenMC and WIMS with ENDF/B-VIII nuclear data library

Jan Malec1, Dušan Čalič1, Andrej Trkov2

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

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

jan.malec@ijs.si

Core design calculations based on deterministic methods can be performed in multiple stages. The CORD-2 cord design package for pressurized water reactors is using three stages, first of which is a fuel cell calculation in a 3x3 lattice with periodic boundary conditions, the second step is a fuel assembly level calculation and the third step a global calculation of the entire reactor core. The first step is performed using WIMSD-5 diffusion code, which calculates homogenized cross sections for the central cell in a 3x3 geometry and the burnup dependent isotopic composition for one fuel cycle. The burnup dependant isotopic composition of the fuel cell is exported to a text-based isolib library, which is interpolated by the CORD2 package to estimate fuel composition at various conditions, such as enrichment, burnup and fuel temperature during burnup. Extensive research has been done in the past on the possibility of replacing the cluster calculation with a stochastic calculation using the Serpent code and delivered promising results despite significantly longer calculation times.
This article describes an inter-code comparison of three cluster cell calculations in a 3x3 lattice with the purpose of comparing not just differences in the results obtained with a stochastic and a deterministic method, but also between multiple stochastic methods. The calculations are performed using the ENDF/B-VIII library, which was prepared using a procedure based on the WLUP project. The newly prepared cross section library includes new fission products, which help narrow down the differences between deterministic and stochastic results.
The comparison focuses on the integral quantity k_inf of the central cell. This quantity is calculated directly by WIMS but needed to be estimated from homogenized cross sections for the central cell using a two-group formula for k_inf derived from the two-group diffusion approximation for both stochastic calculations. Since the k_inf computed this way is not exactly equal to the k_eff computed directly by the Monte Carlo transport solvers, another test case was prepared with just one cell and periodic boundary conditions for verification of the calculation approach based on the two-group approximation.

08.09.2021 10:10 Poster session GREEN

Severe accidents - 401

Simulation of premixed layer formation in PULiMS E6 and SES S1 experimental tests

Janez Kokalj, Mitja Uršič, Matjaž Leskovar

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

janez.kokalj@ijs.si

A hypothetical severe accident in a nuclear power plant can lead to significant core damage, including melting of the core. The hot melt in contact with the coolant water can result in a vapour explosion. Similar explosion phenomenon can be a threat also in some industrial processes, such as foundries and liquefied natural gas operations or in certain volcanic activity where water is present. In analyses of severe accidents in nuclear power plants, a fuel-coolant interaction was mostly addressed in a geometry of a melt jet poured into a coolant pool. Based on some experimental and analytical work from the past a geometry with a continuous layer of melt under a layer of water, called 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) with corium simulants materials contradict this hypothesis. In some of the tests, a premixing layer of ejected melt drops in water was clearly visible and was followed by strong spontaneous vapour 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, developed based on the visual observations and some available mechanisms from the literature, will be presented. The developed model was implemented into the MC3D code (IRSN, France) as a patch and validated against the experimental results of the PULiMS E6 and the SES S1 experimental tests. Analyses of the premixed layer formation parameters will be presented and discussed. The presented analysis will demonstrate the model’s capability to describe the premixed layer formation in agreement with the experimental data.

08.09.2021 10:10 Poster session GREEN

Severe accidents - 412

EVALUATION OF MELT-WATER PREMIXTURE FORMATION DUE TO HYDRODYNAMIC INSTABILITIES

Darya Finoshkina1, Vladimir I. Melikhov2, Oleg I. Melikhov2

1National Research University "Moscow Power Engineering Institute", Krasnokazarmennaya 14, 111250 Moscow, Russian Federation

2Electrogorsk Research & Engineering Centre on Nuclear Plants Safety, St. Constantine 6, 142530 Elektrogorsk, Russian Federation

dfinosh@gmail.com

Hydrodynamic instabilities are important in the formation of melt-water mixture, in which steam explosions can occur. The development of new designs for nuclear power plants and new safety systems is accompanied by an analysis of new situations in which steam explosions can occur. In particular, design-basis accidents with rupture of steam generator tubes can occur in lead-cooled reactors. In this case, high-pressure water flows out as a jet into molten lead, and conditions are created for the occurrence of steam explosions.
Another case is when, in a severe accident at NPP with PWR, the melt falls into a catcher and is filled with water. As a rule, all these events are accompanied by an increase in instabilities at the interphase surfaces, leading to the formation of melt-water mixture.
Recent experiments with high-temperature melts (up to 1400 °C) and water under atmospheric pressure in a stratified configuration revealed strong steam explosions [1]. In this study, a hypothesis was proposed about a new mechanism for mixing a melt with water in a stratified geometry: a thin unstable steam film is formed at the interface between them, from which steam bubbles are formed and condense in subcooled water. With their rapid collapse, high-speed cumulative water jets are developed, directed towards the melt, which produce the splashes of the melt. As a result, a lot of melt droplets appear in the water simultaneously, forming melt-water mixture, in which a strong steam explosion can occur. This hypothesis was proved by numerical calculations in [2]. The initial size of the formed bubbles significantly affects the parameters of the mixture. In this paper, we estimate one on the basis of the Rayleigh – Taylor instability theory.
The stability of the three-layer system "melt (bottom) -steam-water" was analyzed. The potential flow of incompressible fluids was considered. In the frame of linear approach, the developments of small harmonic perturbations of interfaces were analyzed on the basis of a system of linearized equations of hydrodynamics. The kinematic and dynamic conditions for solutions matching on two interfaces made it possible to obtain a system of linear equations, from which the dispersion equation connecting the decrement of the growth of the perturbation with the wave number of the harmonic was derived.
The study of the dispersion relation made it possible to determine the length of the fastest growing harmonic and to estimate the size of the formed steam bubble. For the conditions of experimental studies [1], the calculated bubble size was 1–2 cm, which is consistent with experimental data. As follows from [2], bubbles of this size, when they collapse in subcooled water, are capable of raising up splashes of the melt to a height of 5-7 cm, which is enough to form a melt-water mixture capable of producing strong steam explosions.
A parallel study analyzed the development of surface instability of a high-pressure water jet in molten lead. Using the method of linear analysis of small surface perturbations, the characteristic scales of the formed dispersed particles were determined and the parameters of the formed mixture were estimated.
Acknowledgements. This research was funded by Russian Science Foundation (RSF) under Grant 21-19-00709.
Literature
1. Kudinov P., Grishchenko D., Konovalenko A., Karbojian A. Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials // Nuclear Engineering and Design. 2017. Vol. 314. P. 182-197.
2. Melikhov V.I., .Melikhov O.I., Yakush S.E. and Le T.C. Evaluation of energy and impulse generated by superheated steam bubble collapse in subcooled water // Nuclear Engineering and Design. 2020. Vol. 366. 110753.

08.09.2021 10:10 Poster session GREEN

Severe accidents - 417

INFLUENCE OF NON-CONDENSABLE GAS ON MELT-WATER PREMIXING IN A STRATIFIED STEAM EXPLOSION

Darya Finoshkina1, Vladimir I. Melikhov2, Oleg I. Melikhov2, Sergey E. Yakush3

1National Research University "Moscow Power Engineering Institute", Krasnokazarmennaya 14, 111250 Moscow, Russian Federation

2Electrogorsk Research & Engineering Centre on Nuclear Plants Safety, St. Constantine 6, 142530 Elektrogorsk, Russian Federation

3Institute for Problems in Mechanics Russian Academy of Sciences (IPM RAS), pr. Vernadskogo 101, Moscow, Russian Federation

dfinosh@gmail.com

Severe accidents at nuclear power plants with reactor core melting can be accompanied by steam explosions when the molten core materials come into direct contact with water. Sometimes such a contact can be realized with a stratified arrangement of melt (bottom) and water (top). For a long time, it was assumed [1] that such explosions have a low conversion ratio and do not pose a threat to the integrity of the reactor containment. However, recent experiments [2] have demonstrated that the interaction of a high-temperature melt (1400°C) with subcooled water in a stratified configuration results in strong spontaneous steam explosions with pressures up to 40 bar. Such explosions can take place only in the case of premixing of a significant amount of the melt with water.
In [2], a hypothesis was proposed, according to which such premixing in an initially stratified melt-water system can occur as follows. Steam bubbles form at the interface between the high-temperature melt and water. These bubbles rising in cold water, as a result of which steam condensation begins, and the bubbles collapse. During the collapse, a high-speed cumulative water jet is formed above the collapsing bubble, directed towards the melt. This jet knocks splashes out of the melt. As a result of the superposition of such events, melt-water mixture is formed, capable of producing strong steam explosions.
In work [3], this hypothesis was quantitatively evaluated on the basis of a sequential consideration of bubble collapse and the interaction of the resulting water jet with the melt. It was shown that the arising water impulse acting on the melt surface is sufficient to knock out the melt droplets into the water layer to a height of several centimeters.
Current work is a continuation of the study [3] and is devoted to investigation of influence of non-condensable gas on bubble condensation dynamics in subcooled water. It is assumed that at the initial moment the bubble consists of steam and non-condensable gas. Such a gas can be in water and, when a bubble forms at the interface between the melt and water, it can be in the bubble.
The presence of non-condensable gas in the bubble reduces the rate of bubble condensation. At the same time, at the initial moment, the bubble has a high temperature due to contact with the melt. The hot bubble begins to quickly transfer its heat to subcooled water, which leads to a drop in pressure in the bubble and its collapse.
A numerical study of the collapse of a bubble consisting of steam and non-condensable gas in subcooled water was carried out. We used compressible, heat-conducting model for a bubble description and incompressible, heat-conducting model for water description. The basic features of the process were determined and the kinetic energy of water during bubble collapse was calculated. On the basis of these results, the effect of the influence of non-condensable gas on the mixing of the melt with water was estimated.
Acknowledgements. This research was funded by Russian Science Foundation (RSF) under Grant 18-19-00289.
Literature
1. Berthoud G. Vapor Explosions // Annual Review of Fluid Mechanics. 2000. V. 32. P. 573–611.
2. Kudinov P., Grishchenko D., Konovalenko A., Karbojian A. Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials // Nuclear Engineering and Design. 2017. Vol. 314. P. 182-197.
3. Melikhov V.I., .Melikhov O.I., Yakush S.E. and Le T.C. Evaluation of energy and impulse generated by superheated steam bubble collapse in subcooled water // Nuclear Engineering and Design. 2020. Vol. 366. 110753.

08.09.2021 10:10 Poster session GREEN

Severe accidents - 420

NPP Krško Large Break Loss of Coolant Accident using MELCOR Code

Vesna Benčik, Davor Grgić, Siniša Šadek

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

vesna.bencik@fer.hr

NPP Krško (NEK) input deck for severe accident code MELCOR is being developed at Faculty of Electrical Engineering and Computing (FER) Zagreb. MELCOR is fully integrated computer code that models the progression of severe accidents in light water nuclear power plants. Recently, the MELCOR 1.8.6 input deck was converted to MELCOR 2.1 as well as to a new code version MELCOR 2.2. In this paper the results of Large Break Loss of Coolant Accident (LB LOCA) using MELCOR 1.8.6 as well as MELCOR 2.1 and MELCOR 2.2 are presented. Both unmitigated scenario (engineered safety features not available) and design basis (DB) scenario (one train of Engineering Safety Features (ESF) available) have been analyzed.
The postulated accident is initiated as a guillotine break in cold leg 1 (loop with pressurizer) discharging in Steam Generator (SG) 1 compartment. Simultaneously, an artificial valve connecting two previously connected volumes is closed. In the scenario with ESFs available, one high head and one low head safety injection pump with maximum delay (30 seconds) were assumed available. The accumulator in the broken loop was assumed to spill into containment. Transient was simulated for 10000 seconds. Sensitivity analyses were performed for various values of break discharge coefficients (0.4, 0.6, 0.75 and 1.0) in order to find the most adverse scenario. The results for the analysis with ESF available were assessed against 10CFR50.46 criteria with relation to peak cladding temperature (1204oC) and hydrogen mass (1%). For all three MELCOR code versions the satisfactory behavior of ESFs (1 ECCS train and 1 ESF train in containment), both in RCS and in the containment was demonstrated. For unmitigated scenario only short term results up to 10000 s including time of the Lower Head Failure (LHF) and mass of hydrogen in the core were determined.

Keywords: Large Break Loss of Coolant Accident, MELCOR, severe accidents

08.09.2021 10:10 Poster session GREEN

Nuclear power plant operation - 504

The impact of JEK2 on the operation of the Slovenian electrical power system

Aleksandar Momirovski1, Jurij Kurnik2, Robert Bergant2, Bruno Glaser2

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

2GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia

aleksandar.momirovski@eimv.si

Slovenia is deciding on the direction of the future energy development. The latest international commitments accepted by EU and adopted by Slovenia dictate an increase in electrical energy production from clean low-carbon sources without CO2 emissions and an abandonment of fossil energy sources as a basis for the transition to a low-carbon society. A very reasonable solution to facilitate such transition is the further development of electrical energy production capabilities from nuclear power plants (NPP). In that respect construction of a new nuclear power plant JEK2, with installed capacity of about 1100 MWe, is proposed.
The impact of the new NPP JEK2 on the operation of the Slovenian electrical power system mainly depends on the state of the system generation, consumption, network topology and transit power flows between the adjacent ENTSO-E members. A study of the electrical power system operation from a stationary and dynamic point of view with JEK2 being constructed and connected in 2030 has been performed. The stationary N-1 and N-1-1 security analysis, which also includes generation unit tripping and bus tripping, identifies the most critical network operating conditions and the related critical network infrastructure from a stationary aspect. The dynamic analysis then takes the identified most critical conditions as an input and analyses the impact of failure of individual network infrastructure (trips of main production units and critical failures in transmission system) from a dynamic aspect. Dynamic analysis mainly focuses on the angle and voltage stability in the Slovenian electrical power system. Moreover, a transient stability and critical clearing time analysis of the new NPP JEK2 is presented.
The study includes two scenarios. The first one depicts the state of electrical power system in 2030, when the main generation units in Slovenia (NEK 703 MWe, TEŠ6 542 MWe and JEK2 1100 MWe) could be in operation. The second one shows the state in 2043 with NEK and TEŠ6 being shut down. Planned transmission system upgrades and new or improved connections to neighboring countries are also considered. Along the proposed installed capacity of 1100 MWe, also the maximum acceptable capacity is estimated for secure and stable electrical power system operation.
The results of the study show that the connection of JEK2 unit to the Slovenian electricity network is feasible. Moreover, the new unit along to its substantial low-carbon energy production will significantly contribute to the stability and reliability of the Slovenian electrical power system. The new NPP JEK2 will crucially increase the power system inertia and thus improve its electromechanical disturbance immunity. Consequently, the NPP with traditional reliability and inertia will be essential for development of the unpredictable renewable power sources.

08.09.2021 10:10 Poster session GREEN

Thermal-hydraulics & CFD - 609

TRACE simulation of Semiscale S-NC-2 and S-NC-3 tests

Andrej Prošek, Boštjan Končar

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

andrej.prosek@ijs.si

The RELAP5 and TRACE code developers have selected Semiscale natural circulation tests S-NC-2 and S-NC-3 for validation of prediction capability of RELAP5/MOD3.3 and TRACE V5.0 thermal-hydraulics computer codes. In the case of S-NC-2 test, we have found out that although the same experimental data were used for the assessment of both codes, the primary mass inventory values reported in the two code manuals is somewhat different. Therefore, one the goals of the present study is also to assess the impact of different primary mass inventory on the main output parameters, especially mass flow rate. The latest TRACE V5.0 Patch 6 thermal-hydraulic computer code has been used in the study.
The natural circulation experiments were performed in the Semiscale Mod-2A test facility, which is a small-scale model of the primary system of a four-loop pressurized water reactor (PWR). The facility incorporates the major components of a PWR including steam generators, vessel, downcomer, pumps, pressurizer, and loop piping. The selected experimental tests are Semiscale natural circulation tests S-NC-02 and S-NC-03, consisting of several cases. The S-NC-02 experimental test cases were performed at 60 kW (6% of full Semiscale core power). The objective of the steady-state separate effects S-NC-02 natural circulation test is to study the thermal hydraulic response during the three modes of natural circulation: single-phase, two-phase, and reflux. The secondary side conditions are constant, while the mass inventory on the primary side is varied, causing also variation of other observed parameters (mass flow rates in primary loop, hot leg pressure, hot leg temperature and steam generator outlet temperatures). For S-NC-3 test the system mass inventory for the two-phase conditions is based on S-NC-2 natural circulation testing. The S-NC-3 test cases were performed at a core power of 62 kW, by varying steam generator secondary side mass inventory. The objective of the test performed in the beginning of 1980’s was to study the effect of different steam generator secondary conditions on the two-phase natural circulation.
The ASCII input deck for the RELAP5 has been obtained in the frame of RELAP5 code distribution for the auto validation purposes and has been converted to TRACE input deck using Symbolic Nuclear Analysis Package (SNAP). The simulated results of S-NC-2 and S-NC-3 tests showed that the TRACE V5.0 Patch 6 calculation (using the set of primary mass inventories as reported in TRACE V5.0 Patch 1 developmental assessment validation) agrees very well with updated experimental data, while the original data of primary mass inventory (as reported in RELAP5/MOD3.3 developmental assessment) results in poorer agreement with the S-NC-2 test. The correct value of primary mass inventory used in the simulation of S-NC-2 test is therefore of crucial importance for correct simulation of the next S-NC-3 test cases.

08.09.2021 10:10 Poster session GREEN

Thermal-hydraulics & CFD - 612

Na and NaK Laminar and Turbulent Flows and Heat Transfer within Rectangular Minichannel Heat Sinks

Mahyar Pourghasemi, Nima Fathi

University of New Mexico Department of Mechanical Engineering, MSC 01 1150, Albuquerque, USA-New Mexico

nfathi@unm.edu

Miniature heat sinks have been widely used for heat management of compact microelectronic packages in two recent decades. These small-scale heat sinks operate with minimum amount of coolant while they are durable with long service life. However, the liquid coolant flow regime within these small-scale heat sinks is usually restricted to laminar due to large pumping power required to reach turbulent flow at higher Reynolds numbers. Surface area to volume ratio is large in micro-scale heat sinks and this leads to higher friction and higher pressure drops.
Minichannel heat sinks provide relatively large available heat transfer area to volume ratio that enables them to handle high heat dissipation rates. Moreover, Minichannel heat sinks have larger hydraulic diameters than microchannel heat sinks leading to less pressure drop. Therefore, it is feasible to have turbulent flow regime within minichannel heat sinks with much higher local heat transfer rates than laminar flow. Many published research works are available in the literature on Na and NaK flows and heat transfer within big macro-scale heat exchangers. A few works have been reported so far on liquid metals heat transfer within miniature (mini-scale) heat exchangers. Therefore, the main purpose of this research work is to study laminar and Turbulent flows and heat transfer of Na and NaK in rectangular mini-scale heat sinks. Nusselt numbers and pressure drop values within steel minichannel heat sinks are obtained through 3-D numerical simulations at Reynolds number range of 600-20000. Effect of coolant thermophysical properties on heat transfer rates, is assessed by comparing the obtained Nusselt number values of NaK and Na.

08.09.2021 10:10 Poster session GREEN

Thermal-hydraulics & CFD - 615

LOCA plus Loss of One Emergency Core Cooling System Simulated by RELAP5/MOD3.3 Patch 05

Andrej Prošek

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

andrej.prosek@ijs.si

The second generation reactors were designed and built to withstand without loss to the structures, systems, and components necessary to ensure public health and safety during design basis accidents (DBAs). In the transient and accident analysis the effects of single active failures and operator errors were considered. There are also accident sequences that are possible but were judged to be too unlikely and therefore were not fully considered in the design process of second generation reactors. In that sense, they were considered beyond the scope of design basis accidents that a nuclear facility must be designed and built to withstand. They were called beyond design basis accidents (BDBA).
After Fukushima Dai-ichi 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. In the study, the analysis was performed to see if the selected plant design of two loop pressurized water reactor can prevent spectrum of loss of coolant accidents (LOCAs) together with the complete loss of one emergency core cooling function (e.g. high pressure safety injection (HPSI) or low pressure safety injection (LPSI)). The analysed break spectrum ranged from 1.27 cm to 30.48 cm. For each break size two simulations were performed, one without HPSI pumps and one without LPSI pumps, with other safety systems available. For calculations the latest RELAP5/MOD3.3 Patch 5 has been used and the RELAP5 input model of two-loop pressurized water reactor. Besides this for smaller breaks ranging from 1.27 cm to 5.08 cm leading to core heatup (being DEC as new equipment is needed) accident management strategies to depressurize the primary system in order to enable injection of accumulators and LPSI system have been also studied.
The results show that accident management strategies are successful. However, the control of DEC need to be achieved primarily by safety features for DEC. Therefore it can be concluded that having the safety injection pump with sufficient pressure capability is alternative to primary system depressurization strategy for mitigation of smaller break LOCAs in case the HPSI system is lost.

08.09.2021 10:10 Poster session GREEN

Thermal-hydraulics & CFD - 618

Selection of the most suitable cooling technology for the JEK2 project

Klemen Debelak, Aleš Kelhar, Robert Bergant, Bruno Glaser

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

klemen.debelak@gen-energija.si

The existing nuclear power plant in Krško uses once through cooling with combination of cooling cells if needed. Since the cooling capacity of the Sava River is already fully utilized, JEK2 cooling will depend on cooling towers.

The paper presents the basics of cooling nuclear power plants with cooling towers, the presentation of various technologies on the market and development of computational models for evaluating mechanical end natural draft cooling towers.

Dry, wet, and hybrid cooling technologies are presented. Most nuclear facilities today are cooled by once through or wet cooling. Due to the stringent environmental requirements for once through cooling, more and more newly built power plants are cooled by wet cooling towers.

From a multitude choice of cooling technologies, three alternatives were selected that would be most suitable for cooling the JEK2 condenser: a natural draft counterflow cooling tower, an induced mechanical draft counterflow tower and a hybrid tower with wet and dry cooling. Each of these has its advantages and disadvantages.

Based on the input data for JEK2 project, empirical models developed in MS excel were used to evaluate cooling towers for the main cooling water system (CW) and the essential water supply system (SW). The models link the dependence of JEK2 power and the required dimensions of cooling towers.

The paper also compared the operating costs (excluding maintenance costs) for cooling JEK2 with mechanical cells or a natural draft tower.

08.09.2021 10:10 Poster session GREEN

Thermal-hydraulics & CFD - 621

Simulation of heat transfer in multiple impinging jets with scale-adaptive turbulence model

Martin Draksler, Matej Tekavčič, Boštjan Končar

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

martin.draksler@ijs.si

Though the main features of the jet impingement flow can be reproduced by the time-averaged models, the transient phenomena play a decisive role at the heat transfer prediction on the impingement (heated) surface. These have been rather accurately simulated in our previous studies using the Large Eddy Simulation (LES). But, this approach requires large amount of computational resources that are hardly affordable, especially in the case of multiple jets. The need for less expensive, but still transient, turbulence models is therefore obvious.

A relatively good prediction for the first and second order flow statistics and turbulence budgets can be obtained by computationally less demanding unsteady Reynolds averaged Navier-Stokes (URANS) approach. However, URANS simulations in combination with the conventional eddy-viscosity turbulence models (e.g. Shear Stress Transport (SST) model) tends to suppress the flow instabilities in the shear layer of individual jets that govern the unsteady jet dynamics near the heated surface. Flow unsteadiness is better captured with the SST-based Scale-Adaptive Simulation model (SST-SAS model), which has the capability to detect the local flow unsteadiness and adapt the turbulence quantities (turbulence kinetic energy k and eddy viscosity ?) in these regions. This results in local reduction of eddy viscosity and allows the flow instability to develop.

The objective of this study is to analyze and evaluate the predictive capability of SST and SST-SAS models with respect to turbulence and heat transfer prediction. Investigated case considers the configuration with 13 turbulent impinging jets at inlet Reynolds number of around 20000 and nozzle-to-plate distance equal to four. The numerical simulations are conducted with the open-source CFD code OpenFOAM and validated against the well-resolved LES results.

08.09.2021 10:10 Poster session GREEN

Nuclear regulations - 706

Heading Toward Configuration of BEPU Approach into Licensing Process of SMRs

Seyed Ali Hosseini1, Amir Saeed Shirani1, Reza Akbari2, Francesco Saverio D'Auria2

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

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

reza.akbaryj@gmail.com

08.09.2021 10:10 Poster session GREEN

Nuclear fusion and plasma technology - 805

Thermal Hydraulic Performance Analysis of Smooth and Swirl Tube Type Geometries of Plasma Facing Component Using Entropy Generation Approach

Mohit Pramod Sharma1, Vinay Menon2, Samir Khirwadkar2

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

2Institute For Plasma Research , Bhat Village, Near Indira Bridge, 382428 Gandhinagar, India

mohit.sharma@ijs.si

Efficient thermal performance is important in many heat transfer applications such as heat exchangers, nuclear fuel bundle, and plasma-facing component etc. An important parameter in the design of such systems is the heat transfer and pressure drop. The performance of such engineering systems can also be studied thermodynamically, by considering the net entropy generation in the system.
Entropy generation is a measure of the magnitude of the irreversibility present during that process. Irreversibility in the system can be caused by friction, unrestrained expansion of a fluid, heat transfer through a finite temperature difference, and mixing of two different substances. The higher the irreversibility, the higher will be its entropy generation. Therefore, entropy generation can be used as a quantitative measure of irreversibility associated with a process. It is also used to establish criteria for the performance of engineering devices. The thermal system performance can be improved by minimizing the total entropy generation of the convective heat transfer process. Entropy generation analysis can help designers to determine the extent to which each parameter affects the entropy generation and consequently the efficiency of the system. Based on these predictions and economic issues, the designer will make a final decision to choose the most appropriate design parameters.
This project aims to compare the performance of smooth and swirl tape tube geometries of ITER divertor test mock-up and to study the effects of Reynolds number, twisted tape ratio and length to diameter ratio on entropy generation.

08.09.2021 10:10 Poster session GREEN

Nuclear fusion and plasma technology - 808

Tritium measurements by nuclear reaction analysis using 3He beam

Mitja Kelemen1, Sabina Markelj1, Mickael Payet2, Elodie Bernard2, Matej Lipoglavšek1, Aleksandra Cvetinović3, Christian Grisolia4, Primož Pelicon1

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

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

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

4CEA, IRFM,, F-13108 Saint-Paul-Lez-Durance, F-13108 Saint-Paul-Lez-Durance, France

mitja.kelemen@ijs.si

Tritium (T) is a radionuclide very difficult to analyze because of its chemical and physical characteristics. To investigate T in-situ directly on radioactive wastes coming from fusion or fission facilities, robust, sensitive and cost-effective analytical techniques are rather scarce, especially for low level waste management. An easy way to obtain spots of radioactivity in the field of dismantling is to use smears and proportional counters or the Liquid Scintillation Counting technique (LSC). Following advices from Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) working group, innovative procedures for T and other beta emitters are still required for decommissioning.
Ion beam analysis methods are an important tool in determination of hydrogen isotope retention since ion beam analysis, in particular nuclear reaction analysis (NRA) and elastic recoil detection analysis methods (ERDA), are the only methods that can detect low amounts of hydrogen isotopes on a quantitative level.
Due to the few ion beam methods available for T detection we want to develop a new NRA technique for tritium detection via the 3H(3He,d)4He nuclear reaction. This is analogue to the detection of 2H via the 2H(3He,p)4He reaction commonly used in fusion research. Extensive literature search showed that there were only a few measurements done more than sixty years ago using the reaction with 3He. Their aim was mainly to measure the absolute cross section of the reaction. In order to use this reaction as an analysing tool for low level radioactive waste we need the differential cross section at certain angles. Using a thick tritiated W sample, developed especially for this purpose, we measured the NRA signal in the 3He energy range from 0.7 MeV to 5.1 MeV and we have detected reaction products with energies between 6.5 MeV and 9.75 MeV that were not present on a deuterated W sample prepared in the same manner. This NRA signal proved to be due to deuterium and protons coming from the nuclear reaction between tritium and 3He. The detection signal increased with 3He energy up to 3.4 MeV and decreased with energy at the highest beam energies. At higher energies particles from W and target impurities start to disturb the measurement. Their origin might be due to other nuclear reactions, for example between 3He and W or 3He and 13C. Assuming a constant T concentration of 0.35 at. % down to 3.5 µm we deconvoluted a differential cross section for the 3H(3He,d)4He reaction channel at 135°.

08.09.2021 10:10 Poster session GREEN

Nuclear fusion and plasma technology - 811

Determination of uncertainties for activation material irradiation in TT fusion spectra

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

2Culham Centre for Fusion Energy, Abingdon, Oxon, OX14 3DB, United Kingdom

3ENEA Fusion Division, Via E. Fermi 45, 1-00044 Frascati, Rome, Italy

4EUROfusion Consortium, JET, Culham Science Centre, OX14 3DB, Abingdon, United Kingdom

igor.lengar@ijs.si

A particularly interesting fusion reaction is the TT reaction. In contrast to the DD or DT reactions, the T+T reaction consists of two notable channels: (1) T+T › 4He + 2n, (2) T+T › 5He + n › 4He + 2n. The reaction channel (1) has the higher branching ratio and produces a continuum of neutron energies while channel (2) produces a spectrum with a peak. A particular problem when dealing with the TT reaction is the ratio between the individual reaction channels, which is highly dependent on the energy of the reacting tritium ions.
The problem of determining intensity of the second channel peak can be solved with measurements of the ratio in their intensities. In JET such measurements can be performed with exposure of a suitable set of activation foils in the KN2 diagnostics. The criteria for selection of the optimal foil set have been addressed previously [1]. With a tailored set of activation foils the second channel peak, although relatively weak, can be filtered out [1].
In the process it is very important to correctly evaluate all uncertainties. These include the statistical uncertainties due to gamma ray counting of the activation materials after exposure, uncertainties in the calculated neutron spectra and in nuclear data.
In the present work the evaluation of all uncertainties and the consequent uncertainty in determination of the intensity of the TT reaction channel (2) peak, is performed. The analysis is done for an optimized activation material set, irradiated in the neutron spectrum of the KN2 diagnostics at JET, but the methodology can be used in general. The reason behind the choice for the particular material set is explained, the main restriction being the upper limit for the combined mass of all activation foils irradiated simultaneously.

[1] Igor Lengar et al., Activation material selection for multiple foil activation detectors in JET TT campaign, Fus. Eng. Des., 2018, vol. 136, str. 988-992, doi: 10.1016/j.fusengdes.2018.04.052

08.09.2021 10:10 Poster session GREEN

Nuclear fusion and plasma technology - 814

Development of the MELCOR model for the analysis of large helium ingress into the DEMO cryostat

Rok Krpan, Janez Kokalj, Mitja Uršič, Matjaž Leskovar, Boštjan Končar

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

rok.krpan@ijs.si

The purpose of our research is to analyse the loss of cryostat vacuum due to helium (He) ingress. The initiating event considers a failure of cryogenic lines used for cooling of the Toroidal Field (TF) coils, resulting in spillage of large amount of He into the cryostat. It is assumed that the loss of cooling in TF coils rises the magnet temperature due to electrical resistance heating. Several possible scenarios will be defined, differing in the mass of the leaked He. The total He inventory in the DEMO TF coils is estimated at 8 tons [1] and is based on the extrapolation of the total He mass of 5.2 tons in the ITER magnet coils. Since the DEMO cryostat is designed as a thin-walled structure (in ITER the cryostat is self-standing structure), the cryostat boundary fails at the pressure of 1.0 bar with opening of a rupture disk.
A version of the MELCOR code V1.8.6, specially adapted for fusion reactor safety analyses will be used to perform the analysis. MELCOR is an integral engineering-level code that solves conservation equations of mass, energy and momentum for liquid and vapour phase. It includes several models that describe the physics during the considered accident phenomena. The main fusion related modifications of the MELCOR code, relevant for this analysis, include cryogenic He or air as a primary fluid, modifications for water freezing and ice layer formation, enclosure thermal radiation model and others. Input modelling is based on the “control volume” approach that describes the fusion reactor system fluid volumes and solid structures.
The main goal of this study is first, to develop a generic input MELCOR model for the event of gas ingress into the DEMO cryostat and second, to apply it for the specific scenario of He ingress. The input model requires definition of control volumes for the cryostat, the space between the vacuum vessel (VV) and the vacuum vessel thermal shields (VVTS), the interspace between VVTS and the cryostat thermal shields (CTS) that encloses the magnets and probably also the interspace between the cryostat and the bioshield. Each control volume will need to be connected by the flow paths. The heat transfer will be modelled by the heat structures that involve cryostat walls, VV walls, VVTS, CTS, all magnet coils, gravity supports and the bioshield wall.

[1] B. Končar, M. Draksler, M. Tekavčič, Final Report on Deliverable CFX Assessment of Cryostat air/helium Ingress, EFDA_D_2MDH4H, 2019 February.

08.09.2021 10:10 Poster session GREEN

Materials and ageing management - 906

An investigation neutron irradiation effects on the nanocrystalline boron nitride (h-BN) particles using EPR spectroscopy

Elchin M. Huseynov1, Adil Garibov2, Nicat Abasov2

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

elchin.h@yahoo.com

Paramagnetic centers in the nanocrystalline boron nitride (h-BN) particles were comparatively studied before and after neutron irradiation. Electron Paramagnetic Resonance (EPR) spectroscopic analyzes were performed in the broad range of magnetic field from 0.05 - 0.55 T (500 - 5500 Gauss). The range of 0.3270 - 0.3370 T was additionally sweeped, where more paramagnetic centers were observed. The nature of the new paramagnetic centers formed as a result of neutron transmutation in the BN nanoparticles were examined by EPR spectra. Formation mechanism of VB and VN vacancies has been studied in detail by neutron transmutations.
Two strong signal correponds to free electron of g factor in the EPR spectra of BN nanoparticles were observed as a results of the neutron irradiation [1]. It was found that, the concentration of (_6^12)C isotopes increase as a result of the neutron transmutation. Nitrogen vacancies are formed due to the extremely strong interaction of C atoms with N atoms. In this situation, the nitrogen vacancies are the basis of TBC (triple boron centers) and was observed strong signal appropriate to free electron in the EPR spectra. Another EPR signal was observed as from OBC (one boron center) with increasing the concentration of (_5^11)B isotope under the neutron irradiation. Simultaneously, VB vacancies and defects are formed by the different direction of neutron transmutation at realtively high irradiation doses (4×1015 and 2×1016 n•cm-2) and it was observed as symmetrical EPR signal.

1. Elchin M. Huseynov, Tural G. Naghiyev, Adil A. Garibov, et al. "EPR spectroscopy of neutron irradiated nanocrystalline boron nitride (h-BN) particles" Ceramics International 47, 5, 7218-7223, 2021
2. Elchin M. Huseynov, Tural G. Naghiyev, Nijat R. Abasov "The paramagnetic approach of the color-changing of nano BN particle under the neutron irradiation" Article in press, 2021

08.09.2021 10:10 Poster session GREEN

Materials and ageing management - 909

Microstructural indicators for assessment of the radiation ageing of electronic components

1Slovak University of Technology in Bratislava Faculty of Electrical Engineering and Information Technology Institute of Nuclear and Physical Engineering, Ilkovičova 3, 81219 Bratislava, Slovakia

2Slovak University of Technology Faculty of Electrical Engineering and Information Technology Department of Nuclear Physics and Technology, Ilkovičova 1, 812 19 Bratislava, Slovakia

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

4University of Ontario, Faculty of Energy Systems and Nuclear Science, Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario L1H 7K4, Canada

Radiation ageing of semiconductor devices in electronic components is an important factor for the safe long-term operation of many nuclear and space installations. While low-dose long-term irradiation experiments provide realistic data for assessing the radiation hardness of electronic components, they significantly reduce the potential for fast technological development of reliable and radiation-tolerant electronic devices. On the other hand, there is no engineering consensus on how the results of accelerated ageing experiments can be extrapolated to engineering and design of technologies for long-term applications. A deep understanding of the evolution of the microstructure exposed to accelerated radiation tests inevitably requires employing theoretical modelling and experimental characterization methods. This is a very complicated and challenging task due to the limited size sensitivity of the experimental techniques on the one hand and the limited size of the theoretical calculation’s models on the other hand.
Various techniques based on positron annihilation spectroscopy (PAS) proved themselves to be a powerful tool for identifying crystal lattice defects on the atomic scale in a concentration, which is well below the sensitivity of other experimental characterization methods. By employing conventional radioisotope positron sources such as 22Na, the probed volume in macroscopic (mm-scale) semiconductors samples varies from several tens to few hundreds of micrometres. Such region can be effectively modified (aged) by widely accessible ion implanters.
In our research, we propose 5 MeV proton irradiation in a wide range of fluxes and temperatures for an accelerated material ageing experiment, followed by PAS characterization and complementary testing of the electrical properties of the studied materials/components. In this contribution, we review the state-of-art status in the accelerator ageing of electronic components, based also on their functions, and describe some technical aspects of the proposed experiment.

08.09.2021 10:10 Poster session GREEN

Materials and ageing management - 911

Thick protective Fe coatings for PbLi coolant environments

Jan Cizek1, Jakub Klecka2, Lukáš Babka1

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

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

lukasbabka1999@gmail.com

The introduction of generation IV nuclear fission reactors and fusion systems foresees the potential to use liquid metal-based cooling media. Owing to their favorable chemical-physical characteristics, the most promising candidates are heavy liquid metals such as lead-lithium eutectic (PbLi). The usage of PbLi offers several advantages, such as its very high thermal conductivity that can be employed for rapid and efficient heat transfer. However, using such medium triggers the need to prevent degradation (e.g., corrosion) of the structural materials. A potential way could be deposition of protective, long-term stability coatings onto the surfaces. Radio frequency inductively-couple plasma spray (RF-ICP) is an economical process capable of deposition of thick coatings onto large area surfaces, having a good mutual adherence. Unlike the other, air-based thermal spray methods, its controlled-atmosphere chamber further prevents oxidation, thereby allowing to deposit pure metals and alloys. In our study, we have employed RF-ICP to deposit Fe coatings onto stainless steel demonstration workpieces, manifesting suitability of the method for the task. Following the successful optimization, the Fe coatings were deposited onto structural material coupons (ferritic-type 9% Cr ODS steel) that are to be tested in the PbLi environment.

08.09.2021 10:10 Poster session GREEN

Environment and back end of the fuel cycle - 1009

Methodology of Calculating Air Pollution Dispersion with a Lagrangian Particle Model for Assessing the Impact of Atmospheric Emissions on the environment from Krško NPP

Marija Zlata Božnar, Primož Mlakar, Boštjan Grašič, Darko Popović

MEIS storitve za okolje d.o.o., Mali Vrh pri Šmarju 78, 1293 Šmarje-Sap, Slovenia

marija.zlata.boznar@meis.si

Krško Nuclear Power Plant (NEK) is preparing an Environmental Impact Assessment (EIA) for extending NEK's life from 40 to 60 years until 2043.
Impact on the atmosphere and the population due to accidental emissions is possible. For a quantified assessment we selected a design basis accident with loss of coolant and a beyond-design-basis accident with emission through a Passive Containment Filtered Venting System – PCFVS.
The dose is the criterion for the population impact; three key steps must be performed to calculate it:
• Calculating the source term released into the atmosphere;
• Calculating the atmosphere's dispersion capacities upon emission;
• Calculating the dose based on envisaged radionuclide concentrations at each time and location of the domain of influence.
We will show how we calculated the atmosphere's dispersion capacities during an atmospheric emission.
For a general analysis of a hypothetical accidental event to assess the atmosphere's dispersion capacities, we use the concept of dilution coefficients or relative concentrations X/Q.
Relative concentrations have to be calculated separately for each physical source from which the radionuclides are released into the atmosphere. We have to assess the dilution in the atmosphere caused by emissions from:
• Main ventilation;
• PCFVS stack;
• containment leaks (ground-level emissions).
We assessed the relative concentrations for these three sources for an area of 200 km x 200 km with NEK at the center. We performed the final calculation using three-dimensional meteorological variables fields calculated with the WRF numeric weather forecasts model at a fine spatio-temporal resolution over complex terrain. The dispersion assessment was performed with the Lagrangian particle model Spray using three-dimensional fields of meteorological variables.
We will show how the validations were performed for the models WRF and Spray (comparisons of the model results with measurements under conditions that match the controlled experiment).
The WRF model was validated with measurements at the location of NEK, for which we have the vertical wind and temperature profile up to 500 m above the basin floor.
The dispersion model Spray was validated on an internationally validated dataset for testing dispersion models – “Šoštanj-91”, where SO2 from Thermal Power Plant was used as the tracer for a controlled experiment; we have an inventory of meteorological variables at multiple ground-level stations in the basin and on the nearby hills, and a measured vertical wind profile up to 1000 m.
The validation in the vicinity of Šoštanj over a highly complex terrain for a narrower area helped us to assess the accuracy of using the model for a wider area, which contains a highly complex terrain and a few plains in Slovenia and its neighboring countries.
We will show the average results for the vicinity of NEK and for the target area of 200 km x 200 km.
We will explain the methodology for statistically processing the relative concentration results. That includes calculating the moving averages for different time-averaging intervals, and the percentile values for the temporal and spatial dimension.
We will conclude by showing the statistically processed relative concentration results used to calculate population doses.

08.09.2021 10:10 Poster session GREEN

Environment and back end of the fuel cycle - 1012

Neutron Absorber for VVER-1000 Storage, Transport and Final Disposal Facilities

Martin Lovecky, Jiri Zavorka, Kristýna Klímek Gincelová, Jana Jiřičková, Radek Škoda

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

lovecky@rice.zcu.cz

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 concept of inseparable neutron absorber is introduced to achieve fuel reactivity decrease.
Storage, transport and disposal facilities for VVER-1000 nuclear fuel can be modified with the neutron absorber for better nuclear safety and better economics. For selected fuel handling facilities (spent fuel pool, storage and transport spent fuel cask, final disposal cask), both approaches are used independently. In the first part of criticality safety analysis, neutron absorber is used without facility changes to show maximum increase of nuclear safety in reactivity decrease. In the second part, neutron absorber is used with facility changes for improved economics while achieving the same level of nuclear safety, i.e. the same neutron multiplication factor. Improved economics include boron amount reductions, steel thickness reduction and increased facility capacity.

08.09.2021 10:10 Poster session GREEN

Environment and back end of the fuel cycle - 1015

Effect of the ionizing radiation dose on histone H2AX

Jana Skálová1, Tomáš Vlas2, David Mašata3

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

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

3University of West Bohemia, Faculty of Electrical Engineering, Univerzitni 8, 301 00 Pilsen, Czech Republic

jana.skalova99@gmail.com

In case of the radiation dosimeters are not available, biological dosimetry represents an important method to estimate the absorbed dose of the exposed individuals during nuclear events. Nevertheless, the eukaryotic DNA is constantly exposed to exogenous and endogenous factors. Apart from the ionizing radiation, widescale DNA lesions are also induced by other harmful effects. DNA double-strand breaks (DSBs) are the gravest lesion. DSBs provoke an extensive reaction characterized by the expression of the H2AX molecule. The scope of this work is an assessment of a gamma radiation dose-effect on a human body in terms of expression of the H2AX in DNA.

This contribution is focused on the expression scale of the phosphorylated H2AX molecule (gamma-H2AX), which highlights a DNA damage induced by the exposure to gamma radiation. The dependency between the share of gamma-H2AX molecule in an irradiated sample and the dose of radiation was examined. The investigated subjects consist of fourteen samples of uncoagulable blood from healthy donors. The sample of each donor was divided into four test-tubes - a negative control + three levels of gamma radiation (0,5 Gray, 1 Gray, 2 Gray). The irradiation was performed on a medical caesium source "Gammacell® 1000 Elite." The evaluation was based on the method for determining gamma-H2AX after chemical stimulation DNA – extracorporeal photopheresis.

The outcome of this work is the confirmation that the production of this molecule is dependent on the dose of gamma radiation. Owing to gamma-H2AX characteristic, the finding of the relation between the share of gamma-H2AX molecule in a sample and a dose of radiation was statistically confirmed. Accordingly, the H2AX molecule can be considered a reliable specific marker for DNA damage. In the future, this method could find a purpose in practical events, for example, re-determination dose of radiation after nuclear events.

08.09.2021 10:10 Poster session GREEN

Environment and back end of the fuel cycle - 1018

Spent fuel characteristics of various reactor types

Pauli Juutilainen

VTT Technical Research Centre of Finland, Tietotie 3, Espoo, 02044 VTT, Finland

pauli.juutilainen@vtt.fi

The Serpent 2 Monte Carlo has been used to update the spent fuel characterization calculations over the recent years at the VTT Technical Research Centre of Finland. The purpose of the effort is to determine the impact of uncertainties and parametric choices related to the burnup calculation. Additionally, the impact of the fuel and reactor types is a research object, as the current Finnish nuclear fleet consisting of two VVER-440 and two Nordic BWR units, is about to be extended with an EPR unit soon and a VVER-1200 unit somewhat later.

The impact on source term of fuel and reactor type, uncertainties or approximations related to nuclear data and operating history, as well as computational assumptions has been evaluated for these large-scale power reactors. The first results have been published in conference articles and research reports, e.g. [1-2]. The source term in the context refers to the features that are important to enable safe spent fuel management, such as decay heating power, total activity, spontaneous fission rate and reactivity. These features are produced by the burnup calculation, but the actual significance of the results and differences can be only determined in the further analyses based on these results.

These comparison studies are extended to the category of SMRs in the present article, presenting results from Serpent calculations on a small reactor for electricity production and a reactor specified for district heating. The former is based on the publicly available data of NuScale and the latter on the current design of FinDHR [3] being developed at VTT. The smaller size of the reactor cores in these concepts allows them to be calculated with 3D model even with the Monte Carlo method, but the large core results are based on assembly-wise and 2D full-core calculations.

References:
[1] S. Häkkinen, “Impact of Approximations in Operating History Data on Spent Fuel Properties with Serpent 2”, In Proc. 29th International Conference Nuclear Energy for New Europe, Paper No. 1505, Portorož, Slovenia.
[2] P. Juutilainen, “Effect of burnable absorber rods and U-235 enrichment on EPR UO2 fuel assembly source term with Serpent 2”, VTT Research Report, VTT-R-00242-21, 2021.
[3] J. Leppänen et al., “A Finnish District Heating Reactor: Background and General Overview“, In proc. ICONE-28, Virtual Conference, Aug. 4–6, 2021.

08.09.2021 10:10 Poster session GREEN

Education, training and outreach - 1106

Training Lessons Learned from the Covid-19 in the Spanish Nuclear Fleet Training Centers

Nicolas Moyano

Tecnatom, s.a., Avda. Montes de Oca, 1, 28709 San Sebastian de Los, Reyes, Madrid, Spain

nmoyano@tecnatom.es

Spain has seven reactors in operation that produce more than 20% of the country’s electricity demand. The government agreed with the station’s owners to maintain the current fleet in operation for the period from 2027 to 2035 under the condition of periodic reviews and approval by the nuclear regulator.
Tecnatom is the organization that operates the training centers for the entire nuclear fleet providing the training for the departmental positions related with nuclear safety. These positions are operations (licensed and non-licensed), maintenance, radiological protection and chemistry, as well as initial induction training that support refueling outage activities.
Once the Covid-19 confinement was instituted in Spain in March 2020, the stations decided to suspend or postpone any activity considered non-critical in scope. This decision included training and, as a result, instructors and students could not attend scheduled classes.

In January 2020 there were a total of 27 initial training programs scheduled (12 for licensed operators and 15 for technical personnel) in which there were 143 students and 63 instructors and supervisors involved. Additionally, the requalification and refresher training programs included more than 5000 workers and 84 instructors and supervisors.
The training settings that were planned for all these programs (initial and refresher training) were classroom, that represent 80% of the scheduled time, and practical sessions (control room simulator training sessions for the licensed operators and workshops for technical personnel) that represent approximately 20% of the scheduled time. In all cases the implementation of the training programs were planned in the traditional way where students and instructors are together (synchronous).
The training programs for the refueling outages are primarily conducted by e-learning. So, therefore, the implementation of these programs did not have a negative impact during the Covid-19 crisis.
Actions taken during the confinement
Before the Covid-19 crisis and since 2018, Tecnatom has been working in the development of a specific platform called SOUL (Smart Open Universe of Learning) that applies innovations in training. This was very helpful when the Covid-19 confinement was declared and when the in-person classes were suspended.
SOUL allows the training centers to keep scheduled in-person training in distance mode. This was achieved by the integration of videoconferences and by providing instructor tutorials so that the instructors can work in the new training environment. This also included some apps, such as facial recognition, or on-line quizzes, to help instructors maintain student interest in the subject matter.
SOUL has also facilitated the introduction of innovative training techniques and technologies where the instructor and students are not together at the same time (asynchronous) and the student can learn at his/her own pace (individualized learning).

09.09.2021 10:10 Poster session BLUE

Advances in nuclear technology - 206

A New Reactivity Control Approach for Circulating Fuel Reactors

Giulia Merla1, Antonio Cammi2, Stefano Lorenzi2

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

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

stefano.lorenzi@polimi.it

Molten Salt Reactors are nuclear power systems conceived to be working with a circulating salt-mixture, acting simultaneously as fuel and as coolant, providing safety and sustainability enhancements, other than improvements in the fuel cycle closure. The presence of a circulating fuel features not only the possibility of a continuous Fission Products (FPs) removal but also implies the chance to perform tailored intervention of the fuel isotopic composition, adjusting the mixture as needed. These peculiarities prevent the adoption of commonly available simulation tools, which are designed for solid-fuelled reactors and lack in general the possibility to account for mass exchange processes within burn-up calculations.
To overcome these limitations, a Serpent-2 extension is presented to couple depletion calculations and material transfer, thus featuring the capability to simulate FPs removal and continuous composition adjustments for both reactivity and eutectic control in MSR. The reactivity control is implemented with a new strategy allowing the independence from the chosen fuel treatment and a limited impact on both the system mass and the eutectic proportions. The new implemented functionalities are verified and proven to work correctly, while the new reactivity control strategy is compared to an alternative one from previous studies, proving the superior performances of the new proposal.
Additionally, given the important impact of isotopic changes in the fuel composition on the system behaviour, a specific sensitivity analysis is performed. In particular, the evaluation of sensitivity responses of the main kinetic parameters is exploited to further verify the new reactivity control strategy. Sensitivity coefficients are used to build an analytical tool for control strategy comparison, by which the excellent features of the new proposal are confirmed once.

09.09.2021 10:10 Poster session BLUE

Advances in nuclear technology - 209

Small modular reactors in industrial and district heating combined with thermal energy storage system.

Jan Škarohlíd

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

jan.skarohlid@cvut.cz

With increasing number of small modular reactors (SMR) aimed for industrial and district heating, as DHR400 or TEPLATOR®, a thermal energy storage (TES) seems to be an interesting system feature. Nuclear reactors are generally not designed for rapid power changes, as they are usually connected with unwanted phenomena as xenon poisoning or increased mechanical stress to fuel cladding caused by pellet – cladding interaction. Nuclear reactors are generally better to operate at nominal power, what is more or less easy to do with base load electricity production for large electric grid but may cause several issues in generally smaller district heating grids with more variable consumption profile. On the other side are an industrial (stable with better prediction) consumption or district household heating with variable and less predictable consumption profile. Industrial and district heating systems has different requirements for dynamic response varying from immediate, short term full power delivery (e.g. for finishing already started batch of product before forced manufacturing line cool-down) to long term delivery with more relaxed parameters (e.g. for residual area district heating).
In this study we would like to introduce a concept of a thermal energy storage as a dynamic system connecting SMR and district heating, considering different requirements on both sides with different dynamic system response.

09.09.2021 10:10 Poster session BLUE

Reactor physics & research reactors - 310

The Neutronic Simulation of Traveling Wave Reactor (TWR) Core in the Equilibrium State Using Monte Carlo Method

Islamic Azad University, Department of Nuclear Engineering, Science and Research Branch, Hesarek, 1477893855 Tehran, Iran

Travelling Wave Reactor (TWR) has been proposed as an ingenious design for future nuclear power generation and complete utilization of uranium resources. In this strategy a nuclear fission wave propagates slowly through the core for power production and nuclear fuel transmutation. In equilibrium state of the reactor the shape of neutron flux, nuclide densities and power densities remain constant but the burning region moves in axial direction. The feasibility of fission wave initiation and propagation has been proved widely and led to initial efforts for commercialization studies by the TerraPower Company. In equilibrium state, the unique characteristic of this reactor consists of the in situ fissile material production and consumption leads to the operation without reloading or shuffling of the fuel for reasonable long period of time. This characteristic is consistent with plant economy, proliferation concerns and the energy security, without any need for fresh or spent fuel storage at the site. This research examines the possibility of fission wave formation and neutronic analysis of a Gas-Cooled Traveling Wave Reactor in different cycles up to equilibrium state. The MCNPX2.7 Monte Carlo code has been employed for criticality as well as Burn-up analysis. At each cycle, investigations have been performed for axial fresh fuel loading as well as spent fuel discharge. Results showed that by employing special arrays of neutron absorbers, the equilibrium state is attained from third cycle of the reactor operation where the axial shape of neutron flux and power density distribution remains constant.

09.09.2021 10:10 Poster session BLUE

Reactor physics & research reactors - 313

Advanced analyses of alternative fuel for TEPLATOR DEMO

Tomas Peltan, Eva Vilimova, Radek Škoda

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

peltan@kee.zcu.cz

The TEPLATOR DEMO is a new reactor concept for cheap and clean heat production. The TEPLATOR assumes to use irradiated VVER-440 fuel assemblies, which are not burnt up to their regulatory and design limits. Besides the irradiated VVER fuel, the TEPLATOR DEMO can be operated with a natural uranium fuel, which is a benefit of using heavy water as a coolant and as a moderator. This article builds on previous research work, which concerns with developing a suitable fuel geometry made of natural uranium for TEPLATOR. One of the main goals of this paper is to analyse the fuel behaviour during the reactor operation, for instance the research of an axial burnup profile of designed fuel assemblies for the fuel geometries with the most promising neutronic properties. Further, the reactor core power distribution is calculated. In addition, a behaviour of the designed natural uranium fuel during abnormal conditions is examined. All calculation were carried out in Monte Carlo code Serpent.

09.09.2021 10:10 Poster session BLUE

Reactor physics & research reactors - 316

Silicon carbide neutron detector development computational support with MCNP

Andrej Žohar1, Vladimir Radulović1, Luka Snoj1, Robert Bernat2, Luka Bakrač2, Ivana Capan2, Takahiro Makino3

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

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

3National 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

andrej.zohar@ijs.si

Shortages of helium-3 have caused increased research in the field of semiconductor neutron detectors. The E-SiCure project (Engineering Silicon Carbide for Border and Port Security), funded by the NATO Science for Peace and Security Programme, was focused on the developing radiation-hard silicon carbide detectors with neutron converters. 6-LiF and 10-B4C neutron converter materials were used due to the high reaction rate for the production of charged particles, which were then detected by the SiC detector.
In this paper, the computational support for further developing of the SiC neutron detectors will be presented. An analysis of all possible neutron reactions producing charged particles based on the neutron spectra of the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor will be performed. The aim of the analysis is finding new candidate reactions, or isotopes with high reaction rates, which could be used beside the reactions on 6-Li and 10-B in the neutron converter material and can be experimentally tested at the JSI TRIGA research reactor. Additionally, the analysis will also focus on finding reactions sensitive to fast neutrons as the reactions on 6-Li and 10-B are triggered predominantly by thermal neutrons.
The second part of the computational support for the development of the SiC detectors will consist of charged particle transport in the SiC detector itself with the Monte Carlo N-Particle transport code (MCNP) with the aim to calculate the detector response to charged particles for different neutron converter materials. A detailed model of the SiC detector will be made in MCNP, taking into account the detailed SiC detector construction, including the 70 nm thick Nickel contact on top of the detector. The detector response will be determined in the active part of the SiC detector, a 3 mm × 3 mm × 25 µm active volume (epitaxial layer) on top of a SiC substrate. The energies of produced charged particles (electrons, alpha particles, tritons, etc.) will be sampled from the nuclear data libraries. The sensitivity of the detector to X-rays and gamma rays will also be analysed to determine the background signal.

09.09.2021 10:10 Poster session BLUE

Reactor physics & research reactors - 319

Impact of the various nuclear data libraries on the NPP Krško spent fuel characteristics

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

dusan.calic@ijs.si

Since the neutron nuclear data profoundly influence predictions of the nuclear systems behaviour, it is prudent to evaluate their impact on the spent fuel isotopic composition and consequently their impact on the most important spent fuel observables such as decay heat, activity, neutron multiplication factor, and neutron and photon source term. In this paper ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 libraries are considered. A typical NPP Krško fuel assembly is analysed with the Monte Carlo code Serpent2. Burnup of up to 60 GWd/tU and cooling times up to 100 years are considered in the analysis. Comparison of results showed noticeable differences, which should be taken into account in the library selection and in the uncertainty evaluation of the spent fuel characteristics determination.

09.09.2021 10:10 Poster session BLUE

Reactor physics & research reactors - 322

Impact of Different Fuel Temperature Models on the Nuclear Core Design Predictions of the NPP Krško

Marjan Kromar1, Dušan Čalič2

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

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

marjan.kromar@ijs.si

The CORD-2 system, developed by the Reactor physics department of the “Jožef Stefan” Institute, has been used for the verification of the NPP Krško reload cores since 1990. During recent validation studies, some indications have been noticed questioning the validity of the used fuel temperature model. The model was built based on the calculations with the thermo-mechanical PIN code some 25 years ago. Since than many new fuel performance codes or versions have become available, reflecting the progress in the field. It was decided to revisit CORD-2 fuel temperature model with the results obtained from the FINIX code. FINIX is a fuel behaviour module that calculates the thermal and mechanical behaviour of a nuclear fuel rod during steady-state and transient conditions. It can be used as a standalone code or integrated with the host code on the source level. Since it can be integrated with the Monte Carlo neutron transport code Serpent2, it is especially attractive for coupled temperature – power analysis.
All fuel performance codes require appropriate knowledge of the fuel pin material properties. These properties are usually proprietary and hard or even impossible to get from fuel vendors. Since material properties dictates processes in the fuel rod, such as cladding creep, pellet swelling, densification etc., which are crucial for the fuel temperature determination, lack of material data inevitably reflects into fuel temperature uncertainty. In this paper, assessed variations in material properties are transformed into fuel temperature range over the fuel burnout. The effect of the fuel temperature variations on the neutron transport calculations is assessed with the comparison of the CORD-2 predictions and measurements for all 30 completed NPP Krško fuel cycles.

09.09.2021 10:10 Poster session BLUE

Severe accidents - 410

Comparison of pool scrubbing simulations with SCRUPOS experiment

Matic Kunšek, Ivo Kljenak, Leon Cizelj

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

matic.kunsek@ijs.si

During a hypothetical severe accident in a light water reactor nuclear power plant, the fuel could melt and there is a possibility, that some of the radioactive material could be released as particles to the surrounding area. The releases of the radioactive material can be reduced with the application of pool scrubbing, where the release of contaminated gases is filtered through a pool of liquid water. To understand what is happening during pool scrubbing, phenomena at the local scale have to be understood. Specifically, since the gases enter the scrubbing pool as a jet that disperses into bubbles, the behavior of the particle removal from the bubbles is crucial for understanding the pool scrubbing phenomena.
In the proposed paper, the transition of solid particles from gas to liquid phase during the bubble rise in a scrubbing pool was simulated using subgrid modeling. The multi-phase simulations were performed using the open-source Computational Fluid Dynamics code OpenFoam, with the solver reactingMultiphaseEulerFoam. In the simulations, the gaseous, liquid and two particle phases (phase 1 within bubbles and phase 2 within liquid) were studied. All phases were described in the Eulerian frame. The particle densities and bubble diameters were prescribed, based on data from the literature. The subgrid model takes into account that, due to bubbles rising, the inner air motion moves particles inside bubbles (particle phase 1) due to interfacial drag. The particles first migrate towards the bubble surface and then out of the bubbles. The particles transport from bubbles to liquid is simulated as a transfer via a subgrid model from particle phase 1 to particle phase 2. The subgrid model is programed trough OpenFoam’s flexible framework option called fvOptions, which allows users to add source or sink terms to the differential equations of the OpenFoam solvers. In the end, the results were analyzed and the decontamination factor, which is the resulting measure of the scrubbing efficiency, was calculated and compared with experimental measurements from the SCRUPOS experiment, performed at Politecnico di Milano (Italy).

09.09.2021 10:10 Poster session BLUE

Severe accidents - 413

EMPLOYMENT OF THE ASTEC CODE IN THE SEVERE ACCIDENTS RESEARCH ACTIVITIES AT KIT

Fabrizio Gabrielli1, Victor Hugo Sanchez1, Walter Tromm2

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

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

walter.tromm@kit.edu

The improvement of the performance of the modern integral codes for carrying out severe accident analyses in NPPs is one of the milestones of the Nuclear Safety Research program (NUSAFE) of the Karlsruhe Institute of Technology (KIT). Such activity is performed in the frame of the participation of KIT to European projects and international cooperation on the safety assessment of NPPs including innovative designs, e.g. SMRs, and also aims at supporting decision making under high uncertainties for all emergency situations.
The KIT strategy on the employment of integral codes is based on a clustering the computational issues (development and modelling) with the free access to the in-situ severe accident large infrastructures, e.g. QUENCH, LIVE, HYKA. Such unique research environment, allows performing severe accident research in multiple directions: V&V, development of advanced physical models and mathematical methods, uncertainty quantification, and analysis.
In this framework, the use of the Accident Source Term Evaluation Code (ASTEC), developed by IRSN, plays a central role to analyse the complete SA scenario from the initiating event until radioactive release from the containment in Gen. II and Gen. III water-cooled reactors.
In this paper, the results of recent KIT activities devoted to the ASTEC code validation and to its application for severe accident plant analyses are shown. The QUENCH-08 (PWR bundle) and QUENCH-20 (BWR bundle) are considered for ASTEC validation. The main goal of such tests is analysing the physical phenomena occurring during water injection into a partially degraded core, which is one of the prime SAM measures to prevent the failure of the safety barriers i.e. Reactor Pressure Vessel. In fact, under particular conditions, steam generated during reflooding may significantly enhance the Zircaloy cladding oxidation accompanied by temperature escalation and then trigger a spiky hydrogen generation. The analysis of the results shows that ASTEC is able to reproduce the experimental data of hydrogen production and cladding oxide thickness in instrumented bundle positions. Namely, the code is able to predict the key phenomena governing the effects of reflooding in the early in-vessel phase of a severe accident in PWRs and BWRs.
In line with the KIT strategy for nuclear reactor safety, the plant applications are focused on the evaluation of the Source Term during a severe accident scenario. The ASTEC results of a Medium Break (12’’) Loss of Coolant Accident with and without Station Black Out for a generic KONVOI-1300 NPP are discussed in the paper. The results show the capability of ASTEC to simulate the full scenario, namely simulating the key in-vessel and ex-vessel phenomena as well as their effects on the transport of the Fission Products released by the core to the plant and the environment. In particular, the peculiarity of the ASTEC code to be able to evaluate the element- and isotope-wise composition of the radiological release is of relevance in view of the assessment of a Source Term database to be employed by predictive tools, e.g. JRODOS.
Such validation and analysis activities for ASTEC are therefore of relevance since they integrate the main phenomena affecting a severe accident and the following radiological impact.

09.09.2021 10:10 Poster session BLUE

Severe accidents - 415

Assessment of VVER 1000 SAMG efficiency during LB LOCA accident along with SBO using ASTEC computer code

Pavlin Petkov Groudev, Petya Vryashkova, Antoaneta Stefanova, Rositsa Gencheva

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

antoanet@inrne.bas.bg

This paper presents an investigation of VVER 1000 severe accident management guidelines efficiency during “Large break loss of coolant accident” simultaneously with Station blackout.
The main goal of this assessment has been focused on the investigation of second possible entrance efficiency of KNPP SAMG strategy in case of failure of the first one and an assessment of the possibility to protect the reactor core degradation and the reactor vessel failure. The other goal is an assessment of the main plant parameters behavior, like: core uncovery, core heat up, oxidation of core materials, hydrogen generation, core degradation, fuel cladding failure, partial melting of the core materials with the formation of a molten pool in the reactor core, relocation of core materials to the bottom of the reactor vessel, and formation of a molten pool containing corium. The scenario included quenching of a hot core and recovery of a water level in the reactor core.
In the performed work is included a simulation of operator action, based on a Severe Accident Management Guidelines (SAMG) at VVER1000 of KNPP. The selected scenario is a Large Break Loss of Coolant Accident (LB LOCA) simultaneously with an SBO. The accident starts with a rupture in the cold leg (with ID 850 mm simultaneously with SBO). All HAs have been available during the accident and prevent earlier damaging of the reactor core. The active safety systems are failing because of loss of all AC and DC power sources. In the selected scenario is assumed an operator action based on a SAMG: quenching of the heated core by an injection of borated water in the reactor vessel when the core exit temperature reaches 980 oC by 1 HPP and 1 LPP. It is assumed that the first possible entrance into SAMGs at 650 oC is omitted.
The investigation has been performed with severe accident computer code ASTEC. The activated ASTEC modules of the VVER-1000 input deck are: CESAR, ICARE, SOPHAEROS, RUPUICUV, CORIUM, MEDICIS, DOSE and CPA. All ASTEC modules have been used in a “coupled mode”. The referenced nuclear power plant considered in this investigation is a VVER-1000 reactor of Kozloduy NPP.
The obtained results in this paper could be used for the improvement of severe accident management guidelines (SAMG) as well as for level 2 probabilistic safety analysis (PSA).

09.09.2021 10:10 Poster session BLUE

Severe accidents - 418

Preliminary uncertainty assessment of a severe accident scenario using the ASYST code

Siniša Šadek1, Davor Grgić1, Chris Allison2

1University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia

2Innovative Systems Software, LLC, 1284 South Woodruff, Idaho Falls, Idaho 83404, USA

The ASYST code, developed as part of an international nuclear technology ASYST Development and Training Program (ADTP) managed by Innovative Systems Software (ISS) is an advanced analysis software for nuclear safety applications. It uses best-estimate RELAP5 models and correlations for thermal hydraulic calculation, SCDAP models for calculation of severe accident progression in the reactor core, COUPLE module for the finite element thermal analysis of the reactor vessel lower head filled with molten corium and a variety of other member-developed computational packages. One of these packages is an integrated uncertainty package developed jointly by the Technical University of Catalunya and ISS. The uncertainty analysis enables the user to perform and post-process multiple computer runs to estimate uncertainty bands for desired output parameters. The code generates automatically the file with sampled values of the input parameters, based on the probability density function, the appropriate percentile and confidence level (typically using the default Wilks’ formula).
The uncertainty analysis will be performed for a postulated severe accident scenario of a station blackout at a PWR plant. There are large uncertainties present in the severe accident issues, related to, among others, hydrogen generation during reflood or melt relocation into water, core coolability, in-vessel heat transfer in a damaged core, in-core molten pool behaviour, corium relocation to the lower head, lower head corium behaviour, vessel failure and material release, etc. These issues will be taken into account when determining the appropriate SCDAP parameters to be modified and their probability density functions. The objective of the analysis is to determine uncertainty limits of the most important parameters for assessing the core integrity, the core maximum temperature, primary pressure, hydrogen generation, mass of molten material and the time of the vessel failure.

09.09.2021 10:10 Poster session BLUE

Severe accidents - 422

Overview of experimental data related to melt fragmentation in sodium

Mitja Uršič, Žana Kokot

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

mitja.ursic@ijs.si

The innovative sodium cooled fast reactors could importantly contribute to the sustainability of nuclear power – a low carbon technology. However, for the case of severe accidents with core melt, it is important to estimate the potential of different severe accident management strategies on the reduction of risk for the radioactive release to the environment. One of the foreseen mitigation strategies is to discharge the molten material from the reactor core to the core catcher. During the discharging process the fuel-sodium interaction phenomena will occur.

The melt fragmentation process due to the melt contact with sodium will importantly affect the melt relocation and consequently the fuel-sodium interaction and fuel-structure interaction phenomena loads threatening the integrities of core catcher or reactor vessel.

The aim of our research is to identify and analyse available experimental data to support proposed governing mechanisms of melt fragmentation. The first objective of the paper is to collect experimental data related to the melt fragmentation in sodium. The aim is to collect the data on the melt fragments distribution with morphology, continuous melt break-up length, indications of the hydrodynamic and thermal fragmentation mechanisms. The second objective is to discuss relevant mechanisms for the melt fragmentation.

09.09.2021 10:10 Poster session BLUE

Severe accidents - 423

Analytical Model of Debris Quenching with Top-Flooding Configuration and Additional Gas Injection for FLOAT Experiments

Tim Kelhar1, Markus Petroff2, Janez Kokalj3, Mitja Uršič3, Rudi Kulenovic2, Leon Cizelj3, Jörg Starflinger4

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

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

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

4Institut für Kern und Energietechnik Forschungszentrum Karlsruhe (FZK) GmbH, Hermann-von-Helmholtz-Platz1, Bau 419, D-76344 Eggenstein-Leopoldshafen, Germany

timy.kelhar@gmail.com

A hypothetical severe accident in a nuclear reactor can result in the loss of reactor core integrity. In such case, the aim is to accomplish long-term stable conditions by having a coolable core geometry. Therefore, continuous analytical and experimental research is being performed in this field.
A series of experiments on the FLOAT test facility of IKE, University of Stuttgart, Germany, were performed. The experiments were devoted to study the debris bed coolability in case of top-flooding and counter-current flow of gas (air) injected at the bed’s bottom. The FLOAT test facility complements the DEBRIS test facility by providing the option for simulating a non-condensable gas flow arising from a molten core concrete interaction (MCCI) event.
An analytical model is used to enable preliminary assessment of experimental data from the FLOAT facility, i.e. to perform analysis of quenching behaviour and to determine the impact of MCCI on quenching of debris bed geometry. In the analysed experiments top-flooding configuration was applied with approximately 44 kg of debris (porosity ca. 0.4) and initial debris bed temperatures of 500 °C and 700 °C, respectively. Two out of 4 experiments included gas injection from the bottom.

09.09.2021 10:10 Poster session BLUE

Nuclear power plant operation - 505

The application of machine learning for on-line monitoring Nuclear Power Plant performance

Salvatore Angelo Cancemi, Rosa Lo Frano

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

salvatore.cancemi@phd.unipi.it

The aim of the paper is focused on the development of on-line monitoring strategy and predictive methodology to analyse the performance of the nuclear system and components.
Advancing online monitoring is attracting a lot of interest at nuclear power plants. operating today as it involves the transition from traditional monitoring techniques of nuclear power plant, gathering via manually recorded data sheets, to a full embrace of digitalization.
In this research, a conceptual framework for the application of digital twin technology to primary nuclear power plant component prognosis and maintenance process is proposed in order to reduce its failure risk that could, in turn, affects plant operations and safety.
The development of machine learning algorithm for automated diagnostics and prognostics that, for example, may allow the transition from time-based to condition-based maintenance of the nuclear plant, is totally new and innovative. No prior knowledge of machine learning for on-line monitoring of nuclear items performance in the open literature is known.
The methodology uses big data from sensors and logical controllers for training machine learning algorithm to recognize anomalies or useful pattern before components failure.
Due to the limited available data on primary nuclear components, digital twin concept is adopted in order to generate them for different plant conditions through numerical simulation.
After that, the trained algorithm is capable to predict the performance of nuclear components anticipating or delaying the planned inspection for their repairing/replacing. This approach may support the plant condition-based predictive maintenance optimization and the development of the "digital twin model" for improved plant safety and availability.

09.09.2021 10:10 Poster session BLUE

Thermal-hydraulics & CFD - 607

A neural network model for the microlayer evaporation in wall boiling flows

Ilya Evdokimov, Susann Hänsch

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

i.evdokimov@hzdr.de

Microlayers, the thin layers of liquid forming at the underside of a steam bubble growing at a heated wall, have shown to contribute significantly to the bubble growth under certain wall boiling conditions.

The measurement of their shape and thickness remains an experimental and the simulation of their formation via CFD a computational challenge. Thus, it is difficult to develop microlayer evaporation models covering a wide enough parameter space, which would allow their inclusion into advanced Euler-Euler wall boiling models.

In this work we present a feed-forward neural network (NN), which was trained by a small set of direct numerical simulation (DNS) data with the aim to predict microlayer profiles and volumes under different wall boiling conditions. Various configurations of such machine learning (ML) models were studied and introduced into the OpenFOAM open source CFD solver.

The training data consists of interface-tracking simulation results of the early bubble growth stages. Using the level-set and phase-change capabilities of PHASTA the transient evolution of evaporating microlayer profiles was computed for three different superheats for water at atmospheric pressure. Data mining was then applied to pre-process and feed these results to a neural network in order for it to learn how to predict the microlayer volume depending on different wall superheats and bubble departure sizes.

The computed microlayer-to-bubble volume ratio allowed the trained NN model to be embedded into the RPI wall boiling model of OpenFOAM, which was extended in order to account for an additional microlayer evaporation term. Whilst the overall evaporation component remains unchanged in magnitude, the proposed model does distinguish between the evaporation contributions from the upper curved bubble surface and from the microlayer region.

The NN extended RPI wall boiling model is applied to two demonstration cases: the DEBORA wall boiling case, for which no microlayer contribution is expected, and the experimental case of Lee(2002) for water under atmospheric pressure, for which the microlayer evaporation is expected to be significant. The NN extended RPI wall boiling model is shown to predict reasonable contributions of the different evaporation mechanisms.

The application of ML techniques, where experimental and computational limits hinder sufficient data collection, seems a promising alternative to the conventional development of Euler-Euler models. In the future the NN model presented here can be fed with additional DNS data as well as experimental data for more refined results under various boiling conditions and for different working fluids. The particular implementation of the ML models in the RPI wall boiling model needs to be further researched and discussed with the broad scientific community.

09.09.2021 10:10 Poster session BLUE

Thermal-hydraulics & CFD - 610

Simulations of Containment Atmosphere Mixing with Reduction of Computational Domain

Rok Krpan, Iztok Tiselj, Ivo Kljenak

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

rok.krpan@ijs.si

During a severe accident in a light water nuclear reactor, hydrogen combustion could threaten the integrity of the nuclear power plant (NPP) containment, which could lead to release of radioactive material into the environment. The study of hydrogen distribution in the containment is thus important to predict the occurrence of regions with high local gas concentrations and flammable mixture in order to effectively install mitigation systems in containments.
Various experiments are being performed to simulate atmosphere mixing occurring in NPP containment during severe accidents. During these experiments, a stratified atmosphere is first established and then homogenized by mixing using forced or natural convection. Experiments on mixing and homogenization of a stratified atmosphere are performed in containment experimental facilities of different volumes; from 10 m^3 (small scale facilities) to 200 m^3 (large scale facilities). Within the OECD project SETH-2, experiments of erosion of a helium gas layer with a vertical jet were performed in PANDA experimental facility at the Paul Scherrer Institute (PSI) in Villigen (Switzerland). The PANDA facility, used in this experimental campaign, consists of two cylindrical vessels and an interconnecting pipe, with a total volume of 183 m^3. A computational fluid dynamics (CFD) calculation of single experiment performed in PANDA facility can last up to one month, despite using parallel computing. Furthermore, to perform calculations with needed sensitivity analyses (effect of different numerical meshes, numerical schemes, boundary, and initial conditions) several calculations are needed.
To reduce computational time, computational domains used in calculation can be reduced. In our work, the suitability of reduction methods is studied on experiments performed in the PANDA experimental facility. First, parts of the experimental vessel, which do not contribute much to the mixing process, are not considered in the numerical model. In addition, these experiments are performed within cylindrical vessel with injection taking place in the axis of the vessel or near the vessel wall. In such setup, symmetry or even axisymmetry can be specified. Therefore, simplified axisymmetric two-dimensional or symmetric three-dimensional numerical models of the PANDA facility are developed. The results obtained with reduced models are compared to results obtained with the numerical model of the entire experimental facility and also to experimental results. Finally, the effects of reduced computational domain is assessed and the suitability of the reductions is discussed.

09.09.2021 10:10 Poster session BLUE

Thermal-hydraulics & CFD - 613

Potential of Serpent-OpenFOAM Coupled Codes for Spent Nuclear Fuel Analysis

Tomáš Kořínek1, Jiří Závorka2, Martin Lovecky2, Radek Skoda2

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

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

tomas.korinek@cvut.cz

Operation of future district heating source Teplator with spent nuclear fuel assemblies requires deep analyses. The present article focus on the potential use of coupled neutronics and thermohydraulics codes for simulations with the spent fuel. The simulations were conducted in the neutronic code Serpent and the open-source CFD package OpenFOAM. Both codes were externally coupled with four steps Serpent-OpenFOAM-Serpent-OpenFOAM. The first step was a precursor simulation in Serpent where heat power for CFD simulation was obtained. The second step was a steady-state simulation in OpenFOAM where temperature and density distributions were further used in the third step conducted in Serpent. The heat power distribution from the third step was further used in the last coupling step. The goal of the four-step coupling was to evaluate differences in temperature and heat power distributions between steps. First tests were conducted on one fuel pin of the VVER440 fuel assembly. Additionally, a final simulation was conducted on a simplified fuel assembly with various fuel pin materials. Turbulence was modeled using two-equation RANS turbulent models k-w SST and k-w TNT.

09.09.2021 10:10 Poster session BLUE

Thermal-hydraulics & CFD - 616

On the Choice of Corresponding Pressures in a Novel Fluid-to-Fluid Similarity Theory for Heat Transfer at Supercritical Pressure

Andrea Pucciarelli1, Sara Kassem2, Walter Ambrosini3

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

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

3Dipartimento d'Ingegneria Civile e Industriale - Universita di Pisa, Largo Lucio Lazzarino, 1, 56122 - Pisa (PI), Italy

andrea.pucciarelli@dici.unipi.it

A novel fluid-to-fluid similarity theory, developed in the past years and recently refined and better assessed, allows for achieving an appropriate scaling of heat transfer phenomena at supercritical pressures with different fluid systems. The theory is a sound contribution to define dimensionless groups constituting the boundary conditions to be imposed in designing similar heat transfer experiments for different fluids at supercritical pressure. This achievement is of great importance for broadening the basis of available data in view of the design of the supercritical water cooled reactors (SCWRs).
Previous publications and forthcoming papers describe different details of the similarity theory, showing its success in front of CFD analyses made by DNS, LES and RANS calculations. The aspect dealt with in the present paper relates to the choice of the most appropriate pressure for the different fluids, a necessary step to be performed, conditioning the accuracy of the obtained similarity in the selected operating conditions. Recipes were suggested in previous works for achieving this result, but a systematic analysis of their consequences and their relation to the classical choices dictated by the corresponding state theory were not yet investigated in detail. This analysis is proposed in the present paper, better clarifying the issue in view of future refinements of the rationale at the basis of the similarity theory.

09.09.2021 10:10 Poster session BLUE

Thermal-hydraulics & CFD - 619

Film boiling simulation around cylinder with ANSYS Fluent

Mihael Boštjan Končar, Matej Tekavčič, Mitja Uršič

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

mbkoncar@gmail.com

Sodium cooled reactors are one of the main candidates for the next generation of fast nuclear reactors. Knowledge of the courses of a hypothetical core melt accident are an important nuclear safety issue. Our focus is on the potential interaction of melt with sodium. In this case the rapid and intense heat transfer interaction between the molten core material and the sodium coolant may lead to a vapour explosion. The phenomena and processes during vapour explosion in sodium are under investigation . Due to the nature of liquid sodium (opaqueness, chemical reactivity) the experimental investigation is rather difficult. Therefore, the numerical studies could provide valuable insight into the heat and mass transfer mechanisms. Nevertheless, each numerical simulation, especially multiphase, has to be validated against experimental data. Vapour explosions are experimentally widely investigated in water, where stronger pressure loads than in sodium can be expected. Water experiments can provide a solid basis for validation of numerical models.
In this study we are focusing on the film boiling heat transfer around the melt fragment traveling through the coolant. Benchmark experiments that can be used to mimic the film boiling conditions around melt fragment were conducted in the TREPAM (CEA, France) apparatus [1]. In these experiments the melt fragment are represented by a heated wire moving at a constant velocity through the pressurised water.
Single fluid modelling approach combined with the volume-of-fluid (VOF) interface tracking method will be used to study the heat and mass transfer around the hot cylinder submerged in the water flow. The steady state simulation applying the Reynolds Averaged Navier-Stokes (RANS) approach will be used to resolve the flow. Different eddy viscosity based turbulence models (k-epsilon model, standard k-omega model and k-omega SST model) will be tested to investigate their effect on the film boiling around the cylinder. The main challenge of this work is to determine an adequate boiling modelling approach. To model the phase change during the boiling process, appropriate thermodynamic properties of water will be included in the ANSYS Fluent code. The simulation results, the Nusselt number in particular, will be validated with the data from TREPAM experiment [1].

[1] Berthoud G., D’ Aillon L. G., “Film boiling heat transfer around a very high temperature thin wire immersed into water at pressure from 1 to 210 bar: Experimental results and analysis”, Int J Thermal Science 48, 2009, pp. 1728-1740

09.09.2021 10:10 Poster session BLUE

Nuclear regulations - 704

European Utility Requirements for New Light Water Reactors

Robert Bergant, Jože Špiler

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

robert.bergant@gen-energija.si

The development of existing nuclear fleet in Europe since its beginning in sixties until recently has been done more or less on a national basis without important collaboration between countries. In 1991, 5 major European electricity producers established an organisation to develop a common set of technical requirements, i.e. European Utility Requirements (EUR) for Light Water Nuclear Power Plants in Europe [1]. The main objectives of the EUR organisation are: the development of standard designs, reducing licensing risks, establishing high safety harmonised objectives and promoting cost-effective design features.

The EUR organisation was established in 1991, while the first EUR document was issued a year later. At that time, the EUR requirements were developed mainly on collaboration of aforementioned founders together with the Electric Power Research Institute (EPRI) Utility Requirements Document (URD) in USA. The organisation is changing all the time, currently (as of May 2021) there are 14 European Utilities that are memebers of the EUR organisation.

The EUR organisation with its EUR requirements has several benefits for the development of the nuclear industry, especially Generation III NPPs, in Europe and worldwide. The first one is to share the knowledge, experience and lessons learned of the main european nuclear utlities and based on that, developing the EUR document with regular revisions. The EUR requirements represent the basis for all technical specifications used in bidding documentation in last 20 years. The second one is their applicability for Vendors and Designers willing to sell and build their NPPs in Europe. The Vendors and Designers take the EUR requirments as one of the most important document when developing their designs. The third important outcome is an EUR design assessment process of all those reactor designs which are in interest for European utilities. Practically all Vendors with large LWR designs were already assessed by EUR Utility members or are in the process of the assessment. When the assessment process is successfully completed, the Vendors are awarded by the EUR certificate.

The paper will present the EUR organisation, EUR document, EUR design assessment process and other activities going on.

REFERENCES
[1] European Utility Requirements for LWR Nuclear Power Plants, Revision E, December 2016.

09.09.2021 10:10 Poster session BLUE

Nuclear fusion and plasma technology - 806

Effect of chromium and yttrium addition on the oxidation resistance of tungsten-tungsten carbide composites

Aljaž Ivekovič1, David Simonič2, Irena Paulin3, Saša Novak4

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

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

3The Institute of Metals and Technology, Lepi pot 11, 1000 Ljubljana, Slovenia

4Jožef Stefan Institute, Department for nanostructured materials, Jožef Stefan International Postgraduate School, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

aljaz.ivekovic@ijs.si

Maintenance of structural integrity under high-heat-flux (HHF) fatigue loads is a critical concern for assuring the reliable HHF performance of a divertor. Tungsten is considered as the material of choice for the divertor application due to its intrinsic thermo-physical properties. However, thermal softening, due to recrystallization at temperatures above 1000 °C, can lead to crack formation at the surface as a result of plastic low cycle fatigue. The reinforcement of tungsten by incorporation of carbide nanoparticles results in a material with a stable microstructure and promising high-temperature behaviour.
As such, W-W2C composite is a promising plasma-facing material, however in analogy to pure W, it can cause a substantial safety issue in a loss-of-coolant accident (LOCA) in combination with air ingress into the plasma vessel, due to the formation and sublimation of volatile neutron activated tungsten oxide. Comparable to a passive safety mechanism in self-passivating tungsten alloys, where a stable chromic oxide scale is formed on the surface acting as a diffusion barrier for oxygen and preventing the formation of tungsten oxide, the addition of Cr and Y to W-W2C composites was investigated. The influence of the type and amount of additive on the phase composition and oxidation stability of W-W2C composites was evaluated. Thermodynamic calculations were performed in order to predict the phase composition development and high temperature stability and compared to the experimental results in the as-sintered state and following oxidation resistance testing.

09.09.2021 10:10 Poster session BLUE

Nuclear fusion and plasma technology - 809

Deuterium retention in displacement damaged tungsten-based samples with tungsten carbide inclusions

Sabina Markelj1, Thomas Schwarz-Selinger2, Petra Jenuš3, Aljaž Ivekovič4, Saša Novak5, Mitja Kelemen1, Esther Punzon Quijorna6, Andreja Šestan7, Anže Abram3

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

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

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

4Institut "Jožef Stefan", Jamova cesta 39, 1000 LJUBLJANA, Slovenia

5Jožef Stefan Institute, Department for nanostructured materials, Jožef Stefan International Postgraduate School, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

6Institut Jožef Stefan, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

7Jožef Stefan Institute, Centre for electron microscopy and microanalysis, Jožef Stefan International Postgraduate School, Jamova cesta 39, 1000 LJUBLJANA, Slovenia

sabina.markelj@ijs.si

Current research focuses on the development of tungsten(W)-based materials. They demonstrate some attractive thermo-physical properties; however, these properties deteriorate under operating conditions. For example, the ductile-to-brittle transition (DBT) in the temperature range around 400 °C as well as the recovery, recrystallisation and excessive grain growth at temperatures above 1000 °C, are obstacles for the use of pure tungsten in the proposed operational window of the DEMO reactor. The currently proposed solutions for the structural stabilisation of tungsten include particle reinforcement by the incorporation of oxide (e.g. Y2O3) or carbide (e.g. TiC, TaC) particles into the W matrix that was demonstrated to improve the material’s mechanical properties to a certain extent. As an alternative, it was recently proposed to reinforce tungsten with W2C particles.

As tritium retention is a critical aspect for a future plasma-facing material here hydrogen isotope retention in such W materials reinforced with W2C particles, which are promising DEMO divertor material candidates, was investigated. Four samples with varying carbon content from 4 to 50 at. % were prepared by field assisted sintering in vacuum at 1900 °C, with a heating rate of 100 °C/min, for 5 min and an applied pressure of 60 MPa. To simulate the displacement damage that neutrons will create in the material during fusion operation, samples were irradiated by 20 MeV W ions with a fluence of 7.8x1017 W/m2 at 300 K. In order to be able to compare deuterium (D) retention on un-irradiated material, we irradiated only half of the sample surface. In this way we assure the same exposure conditions for both un-irradiated and irradiated sides of the samples. After the irradiation samples were exposed for three hours at 370 K to a low-temperature D plasma with a substrate bias of -100 V corresponding to an ion energy of 38 eV/D to decorate the defects. Then D depth profiles were measured on each half to measure the D retention by nuclear reaction analysis utilizing D(3He, p)4He nuclear reaction. After this, samples were exposed to D plasma for another 10 hours at the same conditions as before and D depth profiles were again measured afterwards on both sides. We have observed that after three hours of exposure D did not populate the whole damaged zone. This was then achieved after the additional exposure for 10 hours. The D retention is increased on the irradiated part of the sample by approximately a factor of two as compared to the non-irradiated part of the sample for samples that have different percentages of W2C inclusions. On the other hand, there is no difference observed in D retention on pure WC sample and D retention is substantially smaller. This might indicate that such a material is more resistive to displacement damage meaning that it does not trap deuterium more than un-irradiated sample. Detailed microstructure analysis of the samples are performed and results will be presented.

09.09.2021 10:10 Poster session BLUE

Nuclear fusion and plasma technology - 812

Thermal analysis of actively cooled divertor target element for stellarator W7-X

Luka Selan1, Boštjan Končar1, Jean Boscary2

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

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

bostjan.koncar@ijs.si

Wendelstein W7-X is the largest and most advanced stellarator-type experimental fusion reactor located in Greifswald, Germany. After the successful short-pulse operation with an inertially cooled graphite divertor, the transition to reactor relevant materials for the plasma facing components (PFCs) in W7-X is the next necessary step to provide the proof-of-principle that the stellarator concept can meet the requirements of steady-state fusion power reactor operation.
To be able to operate in a steady-state mode and demonstrate a high-performance fusion power, the vacuum vessel walls need to be equipped with actively cooled plasma facing components, the so-called divertor target elements. The foreseen target element will be actively cooled by water and armoured with tungsten. The design of the target element is dictated by its integration onto the complex shaped vacuum vessel, position of water supply pipes, water pumps and weight. The target elements are designed to operate at long pulse operation ? 30 minutes. The heat loading condition and location on the target elements will depend on the operation scenario. Based on the selected geometry, materials and operating conditions, a detailed thermal analysis of a coupled structure-fluid model of the target element will be performed. An appropriate turbulent model for simulation of turbulence in the coolant flow will be selected and validated, along with the necessary mesh convergence analysis. The expected results are temperature distributions in the target structure and distribution of water flow parameters (temperature, velocity, pressure drop) in the cooling channel. The results will be evaluated with respect to the material and coolant temperature limits. The coupled solid-fluid Computational Fluid Dynamics (CFD) simulations will be performed with the ANSYS code.

09.09.2021 10:10 Poster session BLUE

Nuclear fusion and plasma technology - 815

L2G PFC Heat Loads and Field-line Tracing in the SMITER Framework

Leon Bogdanović1, Gregor Simič1, Leon Kos2

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

2University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta 6, 1000 Ljubljana, Slovenia

Integrated within the SMITER Framework, L2G is a new software module for magnetic field-line tracing and heat loads mapping on tokamak plasma-facing components (PFC). The core of the module for field-line tracing is written in C++ and makes use of specialized libraries such as ALGLIB, RKF45 and Intel Embree for cubic B-splines interpolation of magnetic equilibrium data, numerical integration of field-lines and ray (field-line segments) casting with PFC triangle meshes, respectively. The interface (API and GUI) is written in Python and wraps the C++ core through Cython to give the code C-like overall performance. The GUI provides the input of user-defined parameters and options for running field-line tracing cases. Target and shadow geometries can be imported from the SMITER cases, as well as the magnetic equilibrium files. Field-line tracing is done inversely, i.e. the field-lines start on the target geometry and are traced until an intersection with the adjacent shadow geometry is found or the end of the integration interval is reached. Field-lines with no intersections are considered wetted and heat loads on the starting triangles are computed based on the single exponential plasma profile formula. Other relevant tracing results are also computed such as field-line connection lengths and incident angles. The results can be visualized in the ParaView module of the SMITER Framework. Optionally, field-lines can also be plotted. Further speed-up of field-line tracing can be achieved through OpenMP parallelization. This paper presents benchmarking of the L2G module against existing field-line tracing codes (SMARDDA and PFCFLUX) and provides results of tests on PFC meshes of 10 millions triangles.

09.09.2021 10:10 Poster session BLUE

Materials and ageing management - 904

Pressurized Thermal Shock Preliminary Analyses of a 2-Loop Pressurized Water Reactor under Loss-of-Coolant Accident Scenarios

Oriol Costa Garrido1, Andrej Prošek2, Leon Cizelj2

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

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

oriol.costa@ijs.si

Pressurized thermal shock (PTS) concerns the possibility that a reactor pressure vessel (RPV) experiences a reactor transient that subjects its internal surface containing a crack-like defect to a severe thermal shock (or rapid cool down) in combination with pressure. A loss-of-coolant accident due to breaks in primary piping is a type of transient that may lead to PTS due to the injection of cold emergency water required to flood the reactor core. Under these circumstances, a loss of structural integrity of the RPV could occur due to brittle propagation of the crack through the RPV wall. While this type of events are considered in plant safety analyses, they may become life-limiting as the RPV ages. The dominant ageing mechanism with PTS implications is embrittlement, i.e. a reduction of fracture toughness (or resistance to crack propagation) of the RPV wall material with neutron radiation. Thus, the PTS issue is considered an important cornerstone for the long-term operation of nuclear power plants beyond their design lifetime.
A PTS analysis involves multiple disciplines such as thermo-hydraulics (TH), heat transfer, structural mechanics and material science. The TH analysis of the reactor transient aims at obtaining the temperature, pressure and heat transfer coefficient at the downcomer region surrounding the core. These are the inputs to the subsequent thermo-mechanical analysis to compute the temperatures and stresses in the RPV wall, which, in the fracture mechanics analyses, serve to evaluate the stress intensity factor (SIF) histories of postulated cracks at the inner surface of the RPV. Finally, the distance between the obtained SIF curve and the fracture toughness curve of the RPV wall material dictates the available shift in ductile-to-brittle transition temperature, i.e. the PTS temperature margin.
The paper presents the preliminary PTS analyses for a 2-loop pressurized water reactor (PWR) under loss-of-coolant accidents due to breaks in the primary piping. The TH analyses are performed with the latest Reactor Excursion and Leak Analysis Program (RELAP5) code using an available input deck of a Westinghouse type 2-loop PWR developed for cladding peak temperature studies during eventual accident scenarios. The Fracture Analysis of Vessels, Oak Ridge (FAVOR) computer code is employed for the thermo-mechanical and fracture mechanics analyses of several postulated cracks of different types and sizes. Thus, the aim of the paper is to find out the most adverse conditions, in terms of temperature margin, for the RPV under PTS. The preliminary analyses results suggests that, for the accident scenarios considered, the temperature margin is probably sufficient for the safe long-term operation of the plant.

09.09.2021 10:10 Poster session BLUE

Materials and ageing management - 907

Development of a T-junction Structural Model with Cracks for Efficient Fracture Mechanics 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

oriol.costa@ijs.si

Fracture mechanics simulations of flawed structures may be challenging to perform, especially when these involve interdisciplinary and long-transient phenomena such as the turbulent mixing of fluids at different temperatures. In particular, this phenomenon may potentially induce fatigue damage on the structures surrounding the fluids due to varying thermal stresses that arise at the surface caused by temperature fluctuations. Studying the crack growth behavior in such cases typically requires a high-resolution computational fluid dynamics (CFD) simulation of the turbulent mixing to resolve the fluid phenomenon and the structural temperatures simultaneously. The subsequent thermo-mechanical and fracture mechanics analyses of the cracked structure in contact with the fluid allows obtaining the stress intensity factor (SIF) history that governs the crack growth. The latter is then evaluated with a fatigue analysis that translates the SIF ranges during the transient to crack growth rates and, ultimately, to crack growth times. With the exception of the CFD simulation, this tedious process needs to be repeated over different crack positions, sizes and orientations and could benefit from the development of advanced strategies to simplify the data transfer between computer codes, which potentially employ different structural meshes, as well as to minimize computing times and requirements.
The paper presents the development of fracture mechanics models of a cracked T-junction piping, where the fluid mixing occurs, for studying the crack driving force (also known as SIF) with reasonable computing time and resources. This is achieved by using the results of a detailed CFD simulation and a strategy that employs the Abaqus code and Python scripting to optimize the thermo-mechanical and fracture mechanics analyses of the T-junction with arbitrary finite element meshes. The surface stresses and SIF results obtained with the full T-junction model are compared to those obtained with a simplified model using different analysis approaches, namely a “submodeling” technique and a reduced structural model containing the cracks. The results show that, for the piping configuration in-hand, the submodeling approach is not suitable while the reduced structural model provides reasonable outputs as compared to the full model. Thus, the advanced strategy proposed in the paper will be highly valuable for future crack growth analyses of flawed structures exposed to long-transient phenomena.

09.09.2021 10:10 Poster session BLUE

Materials and ageing management - 910

The prediction of failure of pipelines using artificial neural network

Yassine Chahboub

Bay Zoltan Nonprofit Ltd. for Applied Research, Miskolc, Iglói u. 2, 3519 Miskolc, Hungary

yassine.chahboub@bayzoltan.hu

Nuclear energy is the future, year after year the energy need increases, which means more sources of energy will be needed, and especially clean source of energy,
Many countries decided to have a mixed energy approach based on having many clean sources of energy such as solar energy, wind energy… but still, nuclear energy is the most advantageous source of energy because of its efficiency and high productivity.
To ensure the nuclear safety of the nuclear power plant, it is necessary to make sure that all the parts are working properly and safely, one of these parts is the pipeline.
The pipelines are playing a critical role in nuclear power plants as they are transferring heat, water…,
Unfortunately, there is a fact the voids and cracks are present in all kinds of steel especially during the welding, heat treatment process, etc…, which means that it is important to take into consideration their presence.
Many studies were done to predict the failure of pipelines but a gap was found in the time consumed to predict the failure of the pipeline, and here is come to the purpose of this study, which is the prediction of failure of the pipeline in a short time using artificial neural network based on small scale specimen.
The results show that it is possible to reduce the time consumed during the prediction from 20 days to 6 hours, which will a huge help for the nuclear industry especially during the design phase of the nuclear power plants.

09.09.2021 10:10 Poster session BLUE

Environment and back end of the fuel cycle - 1007

Methodology to calculate radiological impact for NPP Krsko Life time extension environmental impact assessment

Davor Grgić1, Siniša Šadek1, Paulina Dučkić1, Primož Mlakar2, Marija Zlata Božnar2, Boštjan Grašič2, Robi Jalovec3, Rudolf Prosen3

1University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia

2MEIS storitve za okolje d.o.o., Mali Vrh pri Šmarju 78, 1293 Šmarje-Sap, Slovenia

3Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

davor.grgic@fer.hr

Regarding the possibility for accident releases of radioactive material, it is clear that even for events with a very small probability density, the likelihood of such a release is increasing with longer operation time. Historically, in case of NPPs, risk is more related to gaseous accident releases. In case of NPP Krsko, the probability for accident release of radioactive liquid effluents is almost negligible. PSA/PRA analysis of NPP Krsko spent fuel pool, after recent plant’s safety upgrades, gives frequency of radioactive release below 1E-08 / year. The corresponding frequency of release from planned spent fuel dry storage is even lower, and it is eliminated by the cask design.
In estimating possibility for environmental influence of radioactive gas effluents due to plant’s accidents, two basis event were considered, a limiting DBA accident, Large Break LOCA, and BDBA SBO accident without any operator action in first 24 hours. In both cases plant specific primary source term is calculated using ORIGEN-S code for real plant cycles. In case of LBLOCA existing NEK Chapter 15 analysis is upgraded from TID-14844 to NUREG-1465 assumptions and dose calculation was performed using RADTRAD 3.03 model. In case of SBO the same RADTRAD release model was used, but thermal hydraulic conditions were obtained from MELCOR 1.8.6. calculation that included actuation of NEK’s PCFV system. X/Q factors for different release conditions (height, thermal conditions), averaged for selected release intervals and for distances up to 100 km were calculated using detailed, Krsko location specific model, based on Lagrangian particle code (SURFPro/Minerve/SPRAY). The obtained 30-days effective dose (TEDE) at 100 km distance was around 4.e-4 mSv for LOCA and around 1.3e-2 mSv for SBO case. That is far beyond allowed dose values for general population (1 mSv per year). The calculation was repeated for distances up to 400 km using JRodos code, for the same radiological release. The obtained 30-days effective doses at the 400 km were more than 10-times lower compared to doses at the 100-km distance (LOCA 4e-5 mSv, SBO 7e-4 mSv). We can see that the resulting cross border influence is very limited even for extremely low probable NEK accidents with significant airborne emissions.

09.09.2021 10:10 Poster session BLUE

Environment and back end of the fuel cycle - 1010

CO2 Mitigation Using Atomic Power-2025 Deployment

William Ernest Schenewerk, 5060 San Rafael Avenue, Los Angeles, CA 90042-3239, Slovenia

CRISPR gene editing will result in 15 billions because human life expectancy will nearly double. This means world energy will continue increasing 2.25%/a through 2100, requiring ~40 terawatts-electric average power generation by 2105, equivalent to 120 terawatts-thermal energy. Sufficient utility-scale energy storage to average 40 terawatts wind and solar energy, ~2 terawatt-a, costing ~2000 trillion USD at 0.10 USD/Wh, will never exist. Wind and solar energy collection cost for ~500 TWe nameplate will add another ~1500 trillion USD, not counting ~500 trillion USD transmission cost. Absent utility-scale energy storage, wind, solar and big hydro will never average more than 3 terawatts electric generation. CO2 mitigation requires atomic power expansion 5%/a from 2025 to 50 TWe nameplate. Otherwise fossil fuel depletion achieves ~1300 ppm CO2 by 2100. Atomic power can be any combination of: (1) seawater-fueled LWR, (2) FBR, or (3) D2O slow-neutron pile. Sufficient D2O will be available from electrolysis and fuel-cell hydrogen consumption. 10 terawatt continuous nameplate power is will be needed to pump 17,500 km^3/a water south from Canada and Russia. Atmospheric CO2 modeling assumes ocean continues absorbing 1/3 of industrial CO2 emissions. Fossil fuel is modeled as gasoline, C8H18. Maximum CO2 is ~850 ppm around 2110. After fossil fuel is phased out, CO2/CH4 (GHG) atmospheric half-life is estimated 83 a, resulting in 350 ppm CO2 around 2350. Results is modification of 2016 Korean prize-winning paper that used now-obsolete 2020 atomic power deployment.

09.09.2021 10:10 Poster session BLUE

Environment and back end of the fuel cycle - 1013

Uendjitjitavi Karupa, Tekla Mutwamezi

Ministry of Health and Social Services - National Radiation Protection Authority (NRPA), Harvey Street, Ministry of Health and Social Services, ministerial Building, 9000, Windhoek, Namibia

uendjitjitavi.karupa@mhss.gov.na

The National Radiation Protection Authority (NRPA) is established by the Atomic Energy and Radiation Protection Act No. 5 of 2005. The Act and its complimentary regulations provides for the protection of workers, public, patients and the environment against the hazardous effect associated with ionizing radiation. The Act provides the basic pillars to implement the system of regulation by notification, registration, licensing, inspection and enforcement in pursuit to control the import/ export, possession and use of sources of radiation.
To ensure the protection of workers liable to occupational exposure the NRPA national laboratory for individual monitoring renders the services to determine and assess the level of exposure. The environmental laboratory is equipped with analytical instrumentation such as the Gamma and the Alpha spec (not fully operational) used to determine the level of radionuclide in environmental samples.

09.09.2021 10:10 Poster session BLUE

Environment and back end of the fuel cycle - 1016

Removal of methyl iodide by self- priming venturi scrubber in a lab scale FCVS setup

1Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 44000, Pakistan

Filtered containment venting system is specifically designed for mitigation of severe accidents. The main objective of this system includes control of over- pressurization by releasing high pressure and temperature gas from the containment. This system ensures removal of radioactive products and refrain them from getting released into the environment. One of the hazardous gases among this list is methyl iodide. An indigenous lab scale setup of FCVS was developed to conduct research on the removal of methyl iodide by self- priming venturi scrubber. Different operational parameters were varied to see their effect on the removal efficiency of methyl iodide. After conducting different experiments, maximum removal efficiency of methyl iodide was obtained which fulfilled the requirement for FCVS.

09.09.2021 10:10 Poster session BLUE

Education, training and outreach - 1104

Remote research reactor exercises during pandemic induced lockdown

Jan Malec1, Vladimir Radulović2, Igor Lengar2, Anže Jazbec3, Sebastjan Rupnik3, Michael Österlund4, Andreas Solders4, Luka Snoj2

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

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

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

4Uppsala University Department of Physics and Astronomy EURATOM-VR Ass., Box 516, SE-75120 Uppsala, Sweden

jan.malec@ijs.si

The Jožef Stefan Institute (JSI) regularly organizes educational courses for local and international students of nuclear engineering and related sciences. The Practical exercises at the JSI TRIGA reactor are highly valuable for students because they allow them to perform hands-on experiments on an operatingnuclear reactor and in some cases the only link between the theory and experimental research. The travel restrictions in 2020 due to the COVID-19 pandemic has prevented international students from travelling to Slovenia and the lockdown in spring 2020, the practical exercises at the reactor were not possible at all.
This paper describes some innovative solutions implemented to allow the participants the closest possible approximation to in-person exercises possible given the restrictions. During the country-wide lockdown in Slovenia, first-ever remote exercises on the Research Reactor Simulator were organized for the students of the Faculty of Mathematics and Physics, University of Ljubljana. While in-person exercises were organized for the same group when the restrictions loosened, the students could perform all exercises planned for the undergraduate reactor physics course on the simulator by following original instructions.
The first remote course was organized for Uppsala University in a five-day practical course in experimental reactor physics. The course was zorganized and performed by using off the shelf but advanced software and hardware components such as: a remotely controlled dome camera in the control room, two conference microphones, portable cameras that the lecturers could take to the reactor, two videoconference setups, remotely controlled laptop used to operate the data acquisition software and the Digital Reactivity Meter, a remotely controlled standard whiteboard, a remotely-operated camera showing the reactor core, cloud document system, videoconference system and remote desktop access system.
The performance of the course was evaluated using an anonymous online survey taken by all the students and their mentor. The organizers provided both open-ended questions and rating scale questions . The aspects being evaluated included the technical content, quality of material, performance of the individual lecturers and the quality of the remote session. In general, the response was overwhelmingly positive, with most questions with a rating scale answered with “excellent”.
In the following months, similar remote exercises were organized in radiation and reactor physics for masters’ students of the University of Ljubljana. In the paper, we describe the education course and its implementation. This is followed by an evaluation of the course and outlook for future improvements.

09.09.2021 10:10 Poster session BLUE

Education, training and outreach - 1107

Augmented Cooperation in Education and Training in Nuclear and Radiochemistry - the A-CINCH Project

Pavel Gabriel Lazaro1, Roberta Cirillo1, Francisco Suarez Ortiz1, Němec Mojmir2, Walther Clemens3

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

2Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Chemistry, Břehová 78/7, místnost: 403, Praha, Czech Republic

3Multiple organizations possible, Unknown, Unknown, Slovenia

gabriel.pavel@enen.eu

The A-CINCH project is the latest project in the series of CINCH projects.
It will primarily addresses the loss of the young generation's interest for nuclear knowledge by focusing on secondary, high school students and teachers and involving them by the “Learn through Play” concept.
This goal will be reached by bringing advanced educational techniques such as state-of the art 3D virtual reality NRC laboratory, Massive Open Online Courses, RoboLab distance operated robotic experiments, Interactive Screen Experiments, NucWik database of teaching materials, or Flipped Classroom, into the NRC education.
Both new tools and existing ones, developed during the previous projects of the series, will be wrapped-up around the A-CINCH HUB – a user-friendly and easy-to-navigate single point of access. The aim is to increase the number of students and trainees in the field of nuclear and radiochemistry.
Nuclear awareness will be further increased by the High School Teaching Package, Summer Schools for high school students, Teach the Teacher package and others.
The main motivation behind this project is that: expertise in nuclear and radiochemistry (NRC) is of strategic relevance in the nuclear energy sector and in many vital applications. The need for radiochemistry expertise will even increase as the focus shifts from safe nuclear power plant operation to decontamination and decommissioning, waste management and environmental monitoring. Applications for NRC are even broader when moving to other fields, they range from life sciences - radiopharmaceuticals, radiological diagnostics and therapy - through dating in geology and archaeology, (nuclear) forensics and safeguards operations, to radiation protection and radioecology.

ACKNOWLEDGMENTS
This project has received funding from the Euratom research and training programme 2019-2020 under grant agreement No 945301.
This project also receives funding from the Norwegian Research Council under grant agreement N° 313053

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#### Dates to Remember

 Abstracts submittal (extended) April 30 May 17 Abstracts acceptance June 30 Early bird registration fee July 31 Young author papers August 10 Full length papers August 23 Conference September 6-9 Proceedings November 30

#### Organiser contact

Društvo jedrskih strokovnjakov Slovenije | Nuclear Society of Slovenia
NENE2021
Jamova cesta 39
SI-1000 Ljubljana, Slovenia
📧 nene2021@ijs.si
📞 +386 1 588 53 31
📠 +386 1 588 53 77
🔗 www.djs.si/nene2021
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