H2020 ARIEL Hands-on school on nuclear data from Research Reactors

25 Sept 2023, 08:00 → 29 Sept 2023, 17:00 Europe/Budapest

Budapest, Hungary

Tamás Belgya (Centre for Energy Research - Budapest Neutron Centre), Szabolcs Czifrus (Budapest University of Technology and Economics)

 

Welcome to the website of the H2020 ARIEL nuclear data school, entitled “Hands-on school on nuclear data from Research Reactors”. 

Date: 25-29 September 2023 

Place: Budapest, Hungary 

Language: English 

Contact: ariel.bpschool@bnc.hu

The registration deadline is now extended until 31 August 2023.

Organizers

The school is organized by the Nuclear Analysis and Radiography Department (NARD) of the Institute for Energy Security and Environmental Safety, Centre for Energy Research (CER), and the Nuclear Technology Institute of the Budapest University of Technology and Economics (BME NTI).

The Centre for Energy Research is the host institute of the Budapest Neutron Centre (BNC) and the operator of the 10-MW Budapest Research Reactor (BRR). BNC manages 16 scientific instruments installed on the neutron beams of the reactor. 

The Nuclear Technology Institute operates a 100-kW Training Reactor.

 

 

 

 

 

 

 

 

 

 

The Budapest Research Reactor at the Centre for Energy Research (left) and the training reactor of the Budapest University of Technology and Economics (right)

About the H2020 ARIEL project 

The EURATOM coordination and support project "Accelerator and Research Reactor Infrastructures for Education and Learning (ARIEL)" brings together the most modern and state-of-the-art European neutron beam laboratories using the full range of neutron sources from high-energy proton synchrotron to research reactors. For the continued improvement of the safety of current and future nuclear facilities, accurate and precise nuclear data are required e.g., for advanced computer simulations. 

Producing these nuclear data is a complex process that relies on neutron facilities and highly trained nuclear physicists. Twenty-three partners from 14 European countries will work together for the education and training of a new generation of young scientists and technical staff.

 

The registration fee and accommodation costs in double rooms will be covered by the H2020 ARIEL Project (www.ariel-h2020.eu). The travel  up to an average quota of about 400 Euro can be covered, which includes fares to come from home to Budapest, back to home and visa if needed.  

It is also recommended to have a health insurance  for the school period.

This school receives funding from the H2020 Euratom research and training program 2014-2018 under grant agreement No 847594 (ARIEL).

 

Budapest_ARIEL_school_poster.pdf

Budapest_ARIEL_school_poster_V03.pdf

Monday, 25 September

08:00 Transport from the Hotel to the campus

Theory: Theory 1

1 Opening session of the school, welcome

Short opening

Speakers: Dr Tamás Belgya (Centre for Energy Research), Dr Máté Szieberth (Budapest University of Technology and Economics), Prof. Attila Aszódi (Budapest University of Technology and Economics)

2 Nuclear data for applications: Evaluation methodology and recent examples

Introduction to the nuclear data & application

Speaker: Dr Roberto Capote Noy (IAEA)

Capote-Budapest-092023.pdf

3 Radiations & their interactions with materials

Speaker: Dr Csaba Sükösd (Budapest University of Technology and Economics)

10:50 Coffee break

Theory: Theory 2

4 Radiative neutron capture cross section measurements in neutron beams

Speaker: Dr László Szentmiklósi (Centre for Energy Research, Budapest Neutron Centre)

Ariel training school SzL 2023.pdf

5 Radiation measurements

Speaker: Dr Imre Szalóki (Budapest University of Technology and Economics)

Radiation measurements-ARIEL.pdf

12:50 Lunch break

Theory: Theory 3

6 Neutron flux measurements with threshold detectors, integral experiments

Speaker: Dr Réka Szilvássy (Budapest University of Technology and Economics)

Neutron flux measurements with threshold detectors_final4.pdf

7 Theory of reactor operation

Speaker: Dr József Kópházi (Budapest University of Technology and Economics)

ARIEL lecture on reactor physics 01.pdf

15:40 Coffee break

Theory: Theory 4

8 Neutron damage of materials

Speakers: Dr Ildikó Szenthe (Centre for Energy Research - Budapest Neutron Centre), Dr András Horváth (Budapest University of Technology and Economics)

Szenthe Horvath ARIEL material ageing 2023 09 25.pdf

9 Neutron-induced fission, prompt gamma-ray measurements, simulation of neutron capture gamma-ray spectra

Speaker: Dr Tamás Belgya (Centre for Energy Research - Budapest Neutron Centre)

ARIEL_school_BT_2023_V01.pdf

Tuesday, 26 September

08:00 Transport from the Hotel to the campus

Flash presentation of the participants: Flash presentations 1 (6 x 15 min)

Milos Travar
Maximilian Kraus
Robert Neagu
Vincent Melzer
Aman Gandhi
Tran Canh Hai Nguyen

10 Short ABSTRACT of Participant's research field

Over the coarse of my faculty studies in the field of physics I was introduced to plethora of different subjects. My background when it comes to Bachelors revolves around Medical Physics. There I was introduced in subjects and techniques that utilize concepts of modern physics and medicine. Here I did research and practical work in the field of RadioTherapy/Diagnostics where my B.Sc. was devoted to optimizing the shielding in the case of PET diagnostics. However, I always showed interest in a more fundamental approach to physics, and during my Masters this interest was recognized by my then mentor and I was employed in CERN within the NICOLE research group (ISOLDE Collaboration). Here we tackled in detail the topic of nuclear structure by exploring certain properties of some exotic nuclei to which my Master Thesis was dedicated. This only further sparked my interests and would send me off to pursue even more ambitious goals towards my future career as a nuclear physics researcher. During my PhD studies I developed high interest in tools for large data analysis such as ROOT, but more importantly I got inspired by the world of Monte Carlo simulations and their applications, which is what I'm predominently working on ever since and which will also be the topic of my Doctoral Dissertation. My efforts in these fields were recognized both by my research group from Faculty and the group I'm working with where I'm employed at the Institute for Physics in Belgrade, where amongst other works we did some novel research in the study of different uranium samples. Some of my international collaborations amongst other include DUNE Collaboration, co-joint research with group for Centre for Energy Research (Budapest) and work at the EU-JRC Institutes with which I honed and improved my skills. Aside efforts at CERN, I've spent time both in JRC-Geel, where I've worked on the study of spontaneous fission of Cf-252 - more precisely investigating the multiplicity of prompt gammas in regards to TKE (but we also tackled some topics concerning neutrons from this process as well), and in JRC-Karlsruhe, where I worked with both Nuclear Forensics Group and the Gamma Spectrometry Team - more precisely we're investigating certain materials and how they could be used as personal dosimeters in case they were to be used as Retrostective Dosimeters, and Monte Carlo modeling of known plutonium samples afterwhich we're aiming at predicting the unknown matrices, which are both considered hot topics in these areas. I would be more than happy to present and expand more in detail my up-to-date work, what my PhD thesis will include and revolve around as well as what plans and motivations I hold for my future career.

Speaker: Milos Travar (Faculty of Sciences, University of Novi Sad, Serbia;)

11 Development of a Time-Dependent Random Ray Method for High-Accuracy Neutron Flux Simulations

The most essential quantity that characterises the behaviour of a nuclear reactor is the neutron flux. Especially for the simulation of its time-dependent behaviour, which is crucial for the safety assessment, relatively coarse simplifications are often applied to make the neutron transport equation accessible to numerical calculations. The presented work introduces a high-fidelity, time-dependent neutron transport solver based on the novel random ray method, a stochastical variation of the method of characteristics. So far, this approach has been tested for a positive reactivity insertion into a homogeneous, infinite model problem, and the obtained results matched those of the analytical solution. The random ray method has a firm potential to provide precise solutions for more complex transients, which shall be investigated in future studies.

Speaker: Maximilian Kraus

12 Four Dimensional Isotope Tracking with the N4DP Instrument at MLZ

Neutron Depth Profiling (NDP) is a non-destructive, element-specific, high-resolution nuclear analytical technique, which is often used to probe concentration profiles of lithium, boron, nitrogen, helium, and several other light elements in different host materials. The N4DP instrument is located at the Prompt Gamma Activation Analysis (PGAA) beamline of Heinz Maier-Leibnitz Zentrum (MLZ), which provides a cold neutron flux up to $5\times10^{10}\,$s$^{-1}$cm$^{-2}$. When a neutron is captured by a specific nuclide, ions with well-defined energies are emitted. The energy loss of the charged particles traveling through the host material is related to the depth of origin at a resolution level up to tens of nanometers.

We developed a detector system based on double-sided silicon strip detectors (DSSSD) with extremely thin and homogeneous entrance windows to provide a new quality of NDP measurements for the N4DP instrument. A highly segmented DSSSD with 32$\times$266 stripes, including integrated, self-triggering electronics, was successfully assessed and tested at the research reactor RID in Delft, in the Netherlands. Using a two-detector camera-obscura geometry set-up, we were able to image and reconstruct Li containing targets with space resolution down to $\sim$100\,$\mu$m$\times$200\,$\mu$m and time resolution down to $10\,$ms. This project is supported by the BMBF, Contract No. 05K16WO1, 05K19WO8.

Speaker: Robert Neagu (TUM)

13 A New Irradiation Workbench Enabling Quantitative Measurements in the Fast and Thermal Neutron and Photon Field of a Polyethylene Shielded Americium-Beryllium Source

The construction of a new irradiation workbench is almost completed. Once finished, it will enable the conduct of quantitative neutron and photon experiments in the field of an americium-beryllium source. Therefore, different radiation field quantities, i.e. spectral fluence rates, fluence rates, ambient dose rate equivalents and directional dose rate equivalents, were determined for the individual fast and thermal neutron and photon components of the source radiation field. This was done by means of evaluated detector measurements and FLUKA simulations at given distances from the source. All results all catalogued in an own database.

The presentation only concentrates on the quantification of the fast neutron field by means of spectral fluence rates. Methods include the calibration of a stilbene detector for neutron energy depositions using a pulse-shape-discrimination-aided Time-of-Flight setup and the unfolding of pulse-charge spectra via a simple unfolding technique. The validity of the results determined by measurements will be discussed by a comparison to analogue FLUKA simulation data.

Speaker: Vincent Melzer (TU Dresden)

14 Nuclear data measurement using accelerator-based neutron sources

In order to develop new applications and advanced technologies the worldwide scientific community requires very precise and highly reliable cross-section measurements. In this talk, I will present a measurement of neutron-induced (n,γ), (n,p), & (n,2n) reaction cross sections in the fast neutron energy region using FOTIA and PURNIMA facilities at BARC, India [1-2]. I will further talk about the work carried out at IFIN-HH related to preparation of a neutron capture cross-section measurement at the n_TOF facility at CERN [3].

References

1. A. Gandhi, Aman Sharma, A. Kumar, Rebecca Pachuau, B. Lalremruata, S.V. Suryanarayana, L.S. Danu, Tarun Patel, Saroj Bishnoi, B.K. Nayak “Neutron radiative capture cross section for sodium with covariance analysis”, European Physical Journal A, 57, 1 (2021).

2. A. Gandhi, Aman Sharma, Rebecca Pachuau, B. Lalremruata, Mayur Mehta, Prashant N Patil, S.V. Suryanarayana, L.S. Danu, B.K. Nayak, A. Kumar “Measurement of (n,γ), (n,p), and (n,2n) reaction cross sections for sodium, potassium, copper, and iodine at neutron energy 14.92±0.02 MeV with covariance analysis", Physical Review C, 102, 014603 (2020).

3. https://www.nipne.ro/proiecte/pn3/ntof/

Speaker: Dr Aman Gandhi (Horia Hulubei National Institute of Physics and Nuclear Engineering—IFIN HH)

15 Comprehensive Review and Application of Neutron Activation Analysis for Elemental Concentration Analysis of Coffee Samples

Neutron Activation Analysis (NAA) is a vital tool in the qualitative and quantitative measurement of elemental concentrations, with far-reaching applications in archaeology, geography, and agriculture, among others. This study applies NAA to quantitatively evaluate elemental concentrations in coffee samples and offers a comprehensive review of NAA — its operational principles and associated methodologies. Utilizing the TRIGA Mark II reactor at the Atominstitut in Vienna and using apple leaves powder samples as a standard reference, we meticulously prepared and irradiated the coffee samples for 180 seconds. The reactor has a power of 250 kW and a maximum thermal neutron flux density of 1x1013cm-2s-2. The samples are fixed in predetermined locations in the reactor core using a pneumatic transport system. Following irradiation, a High Purity Germanium (HPGe) detector was employed to measure the samples. Spectral analysis and comparison of elemental activity enabled the calculation of elemental concentrations. The radionuclides identified are Na, Ca, K, Cl and Mn with masses for each at 1.28 µg, 73 µg, 1349 µg, 10.13 µg and 2.07 µg, respectively. The findings reiterate the effectiveness of NAA in elemental composition quantification, with a spotlight on coffee samples, and underscore the study's contribution to consolidating knowledge around NAA.

Speaker: Ms Tran Canh Hai Nguyen (Dalat Nuclear Research Institute)

10:30 Coffee break

Flash presentation of the participants: Flash presentations 2 (5 x 15 min)

Milos Travar
Maximilian Kraus
Robert Neagu
Vincent Melzer
Aman Gandhi
Tran Canh Hai Nguyen

16 Development of Experimental Methods for Investigating Innovative Approaches to Nuclear Waste Management and to Nuclear Safety

What happens to spent nuclear fuel? One promising future option might be fast (spectrum) reactors operated as a nuclear waste burners, i.e. fissioning plutonium and minor actinides (having half-lifes of tens of thousands of years) and essentially leaving fission products with storage times in the range of hundreds of years. One such fast reactor concept is a molten salt fast reactor (MSFR). The project "NAUTILUS" aims at developing experiments to contribute to the nuclear data base that is needed for the assessment of the parameters and the safety of that reactor concept. The experiments will be conducted at the education and research reactor (AKR-2) of the TU Dresden, Germany.

The three main prerequisites of that goal, which represent major tasks of the project, are (1) the determination of the neutron spectrum of the AKR-2, (2) the simulation of the neutron spectrum of the AKR-2 as well as certain experimental devices, and (3) the development/establishment of the pile-oscillator and the neutron transition method at the AKR-2. The latter methods (3) in combination with the well-characterized neutron field (1, 2) will be employed aiming at reducing uncertainties in the nuclear data base of chlorine-35/37. The gained knowledge will be used to investigate the feasibility of the chloride-based MSFR as a waste burner and assess the safety of the system.

The flash talk motivates the MSFR concept in the given context and gives an overview about the project. Furthermore, the experimental and computational steps gone so far and upcoming work are highlighted.

Speaker: Marco Viebach (TU Dresden)

17 Development and characterization of a novel single plane Compton gamma camera

The detection of radiation in the environment is crucial, often requiring compact and portable instruments. One effective method for detecting gamma-ray sources is through the use of a Compton gamma camera. Unlike Anger cameras, Compton gamma cameras employ source localization techniques based on Compton scattering kinematics. Various types of Compton gamma cameras have been developed, including those based on semiconductor detectors, which offer excellent spatial resolution but limited efficiency and high costs, and those based on scintillator detectors, which utilize either photo-multiplier tubes or Silicon photomultipliers. Scintillator-based Compton gamma cameras generally provide lower angular resolution than their semiconductor-based counterparts, but they offer greater efficiency and cost-effectiveness. Most scintillator-based Compton gamma cameras consist of two separate detector readout planes: the scatterer and the absorber, or they implement complex designs to determine the depth-of-interaction of incoming gamma radiation.

In this study, we introduce a novel concept for a Compton gamma camera, utilizing segmented scintillators read out on a single side by silicon photomultipliers. Each detector element comprises two scintillator crystals optically coupled by a light guide. We employed GAGG:Ce scintillators measuring 3 mm x 3 mm x 3 mm and 3 mm x 3 mm x 20 mm plexiglass light guides. These detector elements were arranged in an 8x8 matrix with a 3.2 mm pitch, separated by ESR reflectors. In this configuration, the front scintillator layer acts as the scatterer, and the back scintillator layer acts as the absorber, both read out by the same silicon photomultiplier array coupled to the back side of the matrix. This distinctive feature minimizes the number of read-out channels, essential for a compact and portable device. The length of the 20 mm light guide was chosen based on a compromise between detector intrinsic efficiency and angular resolution, as determined through GEANT4 simulation studies.

Following the design, we constructed the Compton gamma camera and conducted laboratory tests to characterize its performance. The tests involved Cs-137 and Na-22 sources, with the setup temperature maintained between 18-20°C, and the silicon photomultiplier array read out by the TOFPET2 data acquisition system. Energy resolution characterization was performed individually for the front and back detector layers by irradiating the respective layers with the collimated source. The average energy resolution at 662 keV for the front and back GAGG:Ce layers was found to be 8.9 ± 1.9% and 10.8 ± 1.6%, respectively, accounting for variations within the module.

To analyze the Compton events, we applied specific conditions based on kinematic criteria, such as the energy of each detector element and the geometry of the detector (e.g., distance between fired elements). These conditions ensured a clean sample of Compton events. The basic imaging test using a Cs-137 source (diameter ≈ 3 mm) positioned 50 mm in front of the module, employing a back-projection algorithm, revealed a distinct peak corresponding to the source, with a spatial resolution of σ = 5.1 ± 0.2 mm. In this work, we will present detailed results characterizing the detector's performance, including efficiency estimation and imaging capabilities for gamma sources of different energies at varying distances and positions within the field-of-view. Finally, we will discuss the potential of this designed detector for highly compact and portable Compton gamma camera applications in environmental gamma-ray detection and localization.

Speaker: Mr Om prakash Dash (Department of Physics, Faculty of Science, University of Zagreb)

18 $^{50,53}$Cr (n,$\gamma$) cross section measurement

Chromium is a very relevant element regarding criticality safety in nuclear reactors because its presence in stainless steel, used as structural material. There are serious discrepancies between the different evaluated data of $^{50}$Cr and $^{53}$Cr neutron capture cross section which are not present in the corresponding estimated uncertainties. The Nuclear Energy Agency (NEA) opened an entry in their High Priority Request List (HPRL) to measure these reactions between 1 and 100 keV within 8-10% accuracy. Two experiments have been performed for this matter: one based on the time-of-flight technique at the n_TOF facility of CERN (Geneva, Switzerland) and another based on activation of $^{50}$Cr at the HiSPANoS facility of CNA (Seville, Spain). The experimental set-up and the first results will be presented here.

Speaker: Pablo Perez Maroto (Universidad de Sevilla)

19 NIRR-1'S FUEL ODYSSEY: EXPLORING BURNUP ESTIMATES, DECOMMISSIONING PLANS, AND RADIOACTIVE WASTE MANAGEMENT

This abstract presents a detailed and comprehensive analysis of various aspects related to the Nigerian Research Reactor-1 (NIRR-1), including fuel burnup estimation, decommissioning strategies, radioactive waste management, and proliferation concerns. The study utilizes the WIMS-ANL computer code to assess the fuel burnup characteristics of both the existing Low-Enriched Uranium (LEU) fuel and the decommissioned Highly Enriched Uranium (HEU) fuel in the NIRR-1 reactor. The study focuses on understanding the fluctuation in atom density of key isotopes, namely U235, U238, Xe135, and Pu239, over the burnup time for both the HEU and LEU cores. A comprehensive evaluation of these isotopes' concentrations provides valuable insights into the reactor's operational dynamics and potential safety implications. In terms of fuel consumption, the analysis reveals that during the entire burnup period, the HEU fuel (UAL 4-Al) utilized 9.817257202 grams of U235, while the LEU fuel (UO2) consumed 12.04913749 grams of U235. Similarly, the HEU fuel generated 0.2042625596 grams of Pu239, whereas the LEU fuel produced 0.052503096 grams of Pu239. These findings contribute to a better understanding of the fuel utilization efficiency and the overall performance of the NIRR-1 reactor. The study also highlights the significance of the NIRR-1 reactor as a commercial version of the Miniature Neutron Source Reactor (MNSR) developed by the China Institute of Atomic Energy (CIAE). Operated by the Centre for Energy Research and Training (CERT) at Ahmadu Bello University, Zaria, the NIRR-1 reactor plays a crucial role in neutron activation analysis and minimal radioisotope generation. In line with international efforts to reduce the use of highly enriched uranium in civil nuclear applications, the HEU core of the NIRR-1 reactor has been converted to LEU fuel. Consequently, understanding the current LEU fuel's burnup characteristics is essential for effective fuel management and long-term reactor operation. The analysis confirms the reliability of the WIMS-ANL code in accurately estimating the fuel burnup for the potential LEU core, providing valuable insights for decision-making processes. Furthermore, the study addresses decommissioning strategies for the NIRR-1 reactor, considering its design features and core lifetime. The reactor is designed to operate at maximum flux for 2.5 hours per day, five days per week, with a core lifetime of 10 years. To compensate for the loss of reactivity due to core burnup, controlled beryllium shims are added to the top aluminum tray. This practice leads to increased fuel burnup and the accumulation of fission products throughout the reactor's operation. Considering the potential for proliferation concerns, the study examines the accumulation of fissile materials, particularly Pu-239, in the NIRR-1 reactor. The findings indicate that the accumulation of Pu-239 is insufficient to compensate for the reactivity loss caused by U-235 depletion. Additionally, the concentration of Pu-239 in the spent fuel remains below levels that raise concerns about nuclear weapon design and production. To ensure the safe and efficient operation of the NIRR-1 reactor throughout its life cycle, precise fuel burnup estimation is crucial. This knowledge helps monitor reactivity parameters, neutron fluxes, and power distributions within the reactor core. It also facilitates the estimation of the radioactive source term for safety analysis during accidental scenarios, safeguards requirements for fissile material monitoring, and the determination of cooling and shielding needs for spent fuel storage and transportation. In conclusion, this study provides a comprehensive analysis of fuel burnup estimation, decommissioning strategies, radioactive waste management, and proliferation concerns in the Nigerian Research Reactor-1 (NIRR-1) using the WIMS-ANL code. The findings contribute to a better understanding of the reactor's long-term operation, fuel utilization, and potential safety implications. The results validate the reliability of the WIMS-ANL code for fuel burnup estimation and underscore the importance of accurate fuel management for safe and cost-effective reactor operation.

Keywords: NIRR-1, Fuel odyssey, Burnup estimates, Decommissioning plans, Radioactive waste management

Speaker: Dennis Solomon

20 Efficiency Calibration of a Large Volume Well-Type HPGe Detector

The Canberra GCW6023 well-type HPGe detector is one of the instruments at the Nuclear Security Department that is used to measure the gamma energy spectrum of low activity and typically small-sized environmental samples, e.g., slag, soil, or dust. Small samples in vials can be fitted into the 16 mm wide well providing high efficiency detection and consistent geometry. The detector is also needed for the measurement of other environmental samples which are typically large samples and need to be placed on top of the aluminum cup. Accurate activity determination requires an adequate calibration process for each sample type and sample-detector configuration.
The goal of this project is to determine the full-energy peak efficiency (FEPE) of the GWC6023 detector as a function of gamma-ray energy for different measurement geometries and samples. For this purpose, certified reference materials are assayed, and the Monte Carlo simulation technique is applied to calculate theoretical efficiency values.

Speaker: Judith Dembo

12:15 Lunch break

Theory

21 Demonstration of gamma spectrum evaluation program(s)

Speakers: Dr Tamás Belgya (Centre for Energy Research - Budapest Neutron Centre), László Szentmiklósi (Centre for Energy Research, Budapest Neutron Centre)

22 Introduction to Jupyter Notebooks

Speaker: Dr Gergely Klujber (Budapest University of Technology and Economics)

Jupyter_introduction.pdf

OpenMC_demonstration.pdf

23 Practical demonstration of the use of Nuclear data libraries

Speaker: Dr Oscar Cabellos de Francisco (Polytechnical University of Madrid, Spain)

ARIEL_OCabellos.pdf

Exercise_12-HENDF-JANIS-Wizard-Sep2023.zip

Wednesday, 27 September

08:00 Transport from the Hotel to the campus

Practice @ BME NTI: BME practical session 1

Practice @ the Teaching Reactor

Practice @ BNC: BNC practical session 1

Practice @ BNC

12:00 Lunch break

Practice @ BME NTI: BME practical session 2

Practice @ the Teaching Reactor

Practice @ BNC: BNC practical session 2

Practice @ BNC

Thursday, 28 September

08:00 Transport from the Hotel to the campus

Practice @ BME NTI: BME practical session 3

Practice @ the Teaching Reactor

Practice @ BNC: BNC practical session 3

Practice @ BNC

12:00 Lunch break

Practice @ BME NTI: BME practical session 4

Practice @ the Teaching Reactor

Practice @ BNC: BNC practical session 4

Practice @ BNC

Friday, 29 September

08:00 Transport from the Hotel to the campus

Theory

24 Quiz, Certificate, Student satisfaction questionare, Wrapping up

Speaker: Dr Tamás Belgya (Centre for Energy Research - Budapest Neutron Centre)

Social program