Contents


09.09.2019 17:50 Invited Plenary Speaker 1

Invited lectures – 101

Nuclear non-destructive measurements for nuclear fuel cycle: A review of developments, challenges and prospects

Abdallah Lyoussi

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

abdallah.lyoussi@cea.fr

 

The development of nuclear non-destructive measurement techniques for the inspection, characterisation and analysis of radioactive as well as no radioactive materials began with the birth of nuclear science and technology. Since that time, the control, monitoring and tracking, both of radioactive materials and nuclear facility operations are key aspects that contribute to the quality of scientific and technological programs in the field of physics, nuclear fuel cycle, safeguards and radioactive waste management.
The presentation will start by outlining the history of the French nuclear fuel cycle and the associated measurement, characterization and control evolution and needs. Afterward, a review of the main nuclear non-destructive measurement methods carried out in routine, as well as in particular situations to address specific crucial needs in nuclear fuel cycle operations will be presented. Their advances, challenges and prospects will end the talk.






10.09.2019 08:30 Invited Plenary Speaker 2

Invited lectures – 102

High Fidelity Neutronics Software, uncertainty quantification, and Research Reactors

Alireza Haghighat

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

haghighat@vt.edu

 

Particle transport simulation of nuclear systems is essential for their design, optimization, operation and monitoring. Fast and accurate high fidelity calculation methodologies that are well benchmarked against experimental systems are needed. In the past, i.e., existing reactor technology, simulation tools have relied on somewhat coarse models with approximate methodologies that benefited from two main factors: i) allowance for large margins and tolerances; ii) ability to construct test reactors for adjustment of approximate methodologies. Neither factor can be expected for the development of high fidelity software that are needed for the design, analysis, licensing of the advanced nuclear technologies.
This paper argues the need for new computational paradigms, e.g., the MRT (Multi-stage, Response-function Transport) methodology which has resulted in the development of the novel high-fidelity RAPID (Real-time Analysis for Particle-transport and In-situ Detection) code system. Such code systems have to be robust in modeling any complex system, and should be fast and accurate, henceforth their uncertainties can be quantified at reasonable costs. Due to the lack of access to test reactors, well-characterized and robust research reactors with appropriate environments and high-fidelity, high-precision measurements capability are needed. Such robust facility is offered by the Josef Stefan Institute (JSI) TRIGA reactor and its unique experimental capabilities. Examples of the high fidelity benchmarking studies of RAPID using the JSI’s TRIGA will be discussed.






10.09.2019 14:00 Invited Plenary Speaker 3

Invited lectures – 103

High Fidelity Experimental Data for Computational Fluid Dynamics Validation in Nuclear Applications

Yassin A. Hassan

Department of Nuclear Engineering, Texas A&M University, College Station, Texas 77842-3133, USA

y-hassan@tamu.edu

 

Significant advances have been made in developing computational fluid dynamics tools to simulate the complex flows in nuclear reactor components; however, there remains a need for obtaining high-fidelity experimental data for validation of these computational codes. The transparent linkage between the experiment and the computer program would allow systematic error identification and uncertainty quantification.

Non-invasive methods for flow velocity and temperature measurements have been continuously increasing in popularity and prevalence across many areas of experimental single and multiphase flows and with applications that range across academic research and industrial application. A series of experimental work to achieve high-fidelity measurements of single and multiphase bubbly flows are performed. In this talk several novel experimental techniques aimed at providing experimental databases with high quality, high spatial and temporal resolutions will presented. The talk will cover examples of the results in several practical applications in nuclear reactor applications. The whole-field velocity data are used for validation of computational fluid dynamic computer programs and development of mechanistic models in complex geometries. The use of the experimental data for development of turbulence models such as Reynolds-Average Navier Stokes (RANS), hybrid techniques and Large Eddy simulations (LES) will be discussed.






10.09.2019 16:30 Invited Plenary Speaker 4

Invited lectures – 104

Attracting & Developing New Nuclear Talents: Rebalancing Know-Why and Know-How

Leon Cizelj

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

leon.cizelj@ijs.si

 

Attracting, developing and retaining new talents for nuclear careers are fundamental for the sustainable and safe utilization of nuclear power in the future. Excellent technical specialists understanding the installations of increasing complexity are working in multidisciplinary, multicultural and highly competitive environments.
First signs that the nuclear higher education might be dwindling were noted and reported in high-level documents at the end of the 20th century (e.g., OECD/NEA, 2000, Nuclear Education and Training: A Cause for Concern?). These documents included comprehensive sets of bottom-up and top-down recommendations to preserve and improve the nuclear higher education and training.
Nearly 20 years after the first signs of dwindling nuclear education, the main concerns persist and seem to be reinforced by the phase out of nuclear power declared in many OECD countries. A recent in-depth analysis by Prof. Bum-Jin-Chung (Attracting a High Quality Nuclear Workforce- Recollection of the Nuclear Knowledge Management, keynote lecture at 3rd IAEA Int Conf on Human Resourced Development for Nuclear Power Programs, May 2018, Gyeongju, Korea) pointed to some very plausible reasons for the persistent concerns, including:
• Tendency to solve the easy problems first;
• Tendency to be more concerned about HOW and WHAT then WHY.
An insight in the current nuclear ETKM situation will be discussed and in terms of the inherent incompatibility of the “know-how” and “know-why”cultures.
WHY is usually associated with curiosity, knowledge, higher education, research, and academia. Similarly, HOW and WHAT may be associated with needs, training, skills, experience, knowledge management, industry and knowledge communities.
Some possible ways to overcome the persistent concerns in nuclear education will be proposed also proposed. All of them call for high level of support, coordination and partnership between all nuclear stakeholders, especially those involved in all levels of decision-making.






11.09.2019 08:30 Invited Plenary Speaker 5

Invited lectures – 105

Can Nuclear Power Solve Climate Change?

Alastair C. Laird

European Nuclear Society, Rue Belliard 15-17, B-1040 BRUXELLES, Belgium

alastair@the-lairds.com

 

Business Cases have both a time dimension and a content dimension. The time dimension is specific to maturity – from early strategic thinking (Strategic Case); to technology maturity (down-selection and preference); to the final Case for Financial Investment as the precursor to project delivery. Each time we re-visit the Business Case we seek to answer different questions – against the context of investment and benefit.
a) The Strategic Case – WHY are we doing this and what be different – the outcomes
b) The Economic Case – VALUE of delivering the program in a particular way against alternative options
c) The Financial Case – AFFORDABILITY to do it
d) The Management Case – HOW we can set up national resources up to deliver the right international outcomes
e) The Commercial Case – The case for international market integration, communities of experts and collaboration across scientific boundaries

In many ways our past constrains our future. Site locations, technology selection, vendors and operators – change is not in our DNA!
For civil nuclear power, the argument has also historically been aligned to the economics [€/kWh – electricity prices] of the market against perceived safety dis-benefits. The investment, in all European countries, has unsurprisingly been made within the framework of the national balance sheet. The early NPPs were not private endeavours or market ready projects, these were strategic investments where countries conducted their own R&D, created academic institutes and developed their own intellectual capital. National nuclear utilities were then formed or grew out of existing power utilities. This model is no longer sustainable and the world now has the issue of climate change to tackle. We are in danger of fiddling whist Rome Burns. As for change – what are our options and what is the right answer to the 5 questions (a) to e) above. Ultimately the business case should converge on a single preferred solution and overcome the politic of preference and the market for a fast buck.

Is ITER an example we can learn from?
Is standardisation on reactor type an area we should focus on?
Is the timing now or can we afford to wait another decade or two?






12.09.2019 08:30 Invited Plenary Speaker 6

Invited lectures – 106

The new Divertor Tokamak Test facility

Piero Martin

University of Padova, Via 8 Febbraio 1848, 2, 35122 Padova PD, Italy

piero.martin@igi.cnr.it

 

Appropriate disposal of the non-neutronic energy and particle exhaust in a reactor is universally recognized as one of the high priority challenges for the exploitation of fusion as an energy source. The new Divertor Tokamak Test (DTT) facility , which will be built in Italy, is a tool to address that challenge in high-field, high performance tokamak with complete integration between core and edge plasma scenarios. The Divertor Tokamak Test facility is a superconducting tokamak with 6 T on-axis maximum toroidal magnetic field carrying plasma current up to 5.5 MA in pulses with total length (including start-up, flat-top and ramp-down phases) up to about 100 s. The D-shaped device is up-down symmetric, with major radius R=2.14 m, minor radius a=0.65 m and average triangularity 0.3. The auxiliary heating power coupled to the plasma at maximum performance is 45 MW, shared between the three heating systems used in ITER and foreseen in the present version of DEMO: ion and electron cyclotron resonance heating and negative ion beams. In particular DTT will use 170 GHz ECRH, 60-90 MHz ICRH and 400 MV negative ion beam injectors. DTT will start operation with 8 MW ECRH, which will increase to 25 MW within two years from the beginning of plasma operation. The full 45 MW heating power is planned to be reached within 6 years. An input power of 45 MW allows matching PSEP/R values, where PSEP is the power flowing through the last closed magnetic surface, with those of ITER and DEMO (i.e around 15 MW/m). The plasma facing material is tungsten, sprayed on the first wall and bulk in the divertor. The entire machine is up-down symmetric to allow the study of double null (DN) divertor configurations and will be maintainable through remote handling. The central solenoid, the toroidal and the poloidal field coils will all be superconducting (mostly Nb3Sn), with the possibility of adding in a second phase a high-temperature superconducting insert in the central solenoid to test this emerging technology in a fusion environment. The external coils together with a set of six internal coils will allow to stabilize, control and optimize the local magnetic configuration in the vicinity of the divertor target. The main divertor magnetic topologies, which can be produced in DTT are the reference single null, double null and snowflake configurations. These can be produced at (or close) to the maximum target plasma current of 5.5 MA, while double super-X may be feasible only at significantly lower current (and with vertical stability issues still to be solved). The DTT poloidal field coils system also allows for the realization of scenarios with negative triangularity. In particular a 5 MA single null scenario with ?=-0.13 and ?_lower=-0.16 and a double null scenario at 3.5 MA with ?=-0.38 can be produced. This presentation will discuss the state of the art of the project, illustrating its scientific background, the expected plasma scenarios – in particular as far as plasma exhaust is concerned – and the main technology choices so far. Emphasis will be given to highlight the effort to design an experimental tool, which will be a device not only for plasma exhaust studies, but also for the advancement of fusion science in the grand sense.

1 https://www.dtt-project.enea.it/downloads/DTT_IDR_2019_WEB.pdf






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 204

Simulation of Erosion of a Helium Gas Layer with a Vertical Air Jet in SPARC Test Facility

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, a hydrogen explosion could threaten the integrity of the nuclear power plant containment, which could lead into release of radioactive material into the environment. Various experiments are performed to simulate physical phenomena occurring in containment during severe accidents and results are used to validate Computational Fluid Dynamics (CFD) codes in order to simulate phenomena in actual power plants.
The generation of a stratified atmosphere and mixing of an air-helium mixture was observed in the large-scale test facility named SPARC (Spray-Aerosol-Recombiner-Combustion) at Korea Atomic Energy Research Institute (KAERI) in Daejeon (Korea). The SPARC facility consists of a single cylindrical vessel with a volume of 80 m3.
The interaction of axisymmetric air jet on a horizontal layer of air-helium mixture in the upper part of the vessel was simulated using the open-source CFD code OpenFOAM v1606+. A two-dimensional mesh using a wedge boundary condition and a three-dimensional axisymmetric numerical model of a single cylindrical vessel of the SPARC test facility were developed. Simulation results (local helium concentrations) from both models are compared with experimental results.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 205

Preliminary Assessment on Containment External Cooling Effect for FLEX Strategy using Containment Analysis Codes

Kyungho Nam

Korea Hydro & Nuclear Power Co. Ltd., 70gil, Yuseong-Daero 1312, Yuseong-Gu, 34101, Daejeon,, South Korea

khnpkhnam@khnp.co.kr

 

The external containment cooling strategy is involved in the FLEX Support Guideline (FSG). According to this document, the external containment cooling strategy will be most effective if the steel containmnet vessel itself can be sprayed with cool water. additionally, this cooling strategy should be evaluated for plant-specific containment building design. In case of Korean nuclear power plant, the material of containment building is pre-stressed concrete. Therefore, it should be checked that the external cooling strategy which is specified in FSG has an effect on the depressurazation of containment building. In this paper, the containment external cooling effect was anlyzed using GOTHIC and CAP codes. In order to invertigate the influence of external cooling, an Extended Loss of All AC Power (ELAP) condition which is one of the entry condition for FSG-12 was applied. Additionally, the maximum RCP leakage was also assumed. As a calculation results, it was showed that the external spray cooling effect using portable pump have little effect on depressurization of containment building after 48 hours. And, the containment pressure is low sufficiently to maintain the containment integrity and implement other mitigation strategies.






10.09.2019 15:40 “Poster Session I ” and 11.09.2019 10:10 “Poster Session II “

Thermal-hydraulics – 206

Water jet pump design by using analytical calculation, CFD analysis and model tests

Franci Vehar, Rok Pavlin

Kolektor Turboinštitut d.o.o., Rovsnikova 7, 1210 Ljubljana, Slovenia

franci.vehar@kolektor.com

 

The new built nuclear power plant uses the combination of centrifugal and water jet pumps at the emergency cooling down system. According to the system requests water jet pumps should be designed for three different operating points.

The purchaser had available test results of preliminary designed water jet pump. During the tests of the full scale jet pumps, the requested operating points were not reached due to the cavitation breakdown.

Kolektor Turboinstitut took over the development of a new jet pump. The paper describes the design and production of the jet pumps in following steps:
1. Initial hydraulic design
2. Detailed hydraulic design including CFD simulations and model testing
3. Mechanical design and dimensioning
4. Production of four full scale jet pumps.
At all steps the preliminary design was used for the comparative purpose.

Initial hydraulic design was done by studying different literature and using different design methods: Karrasik, ESDU 85032, Sokolov and Shultz method. Each method uses its own analytical calculation to predict jet pump characteristics and to predict the cavitational behaviour. The main dimensions of the jet pump are included into the characteristics equations.

Detailed hydraulic design including CFD simulations and model testing was performed as the next step. For the flow analysis, the computer code ANSYS-CFX has been used. The governing equations were the Reynolds averaged Navier-Stokes Equations. Either the K-? SST or the EARSM turbulent model were used to provide a link between the turbulent transport of momentum and energy and mean flow properties. During the development, different geometries were analysed, differing some main parameters like diameter, position and shape of the nozzle, shape and diameter of the throat. After finishing the CFD analysis the most promising designs were tested at the model in scale 1:2. The model was made of aluminium (suction chamber), stainless steel (nozzle, diffuser) and plexi glass (throat). There was the possibility to observe the flow through the nozzle through the plexi glass and detect the cavitation level at the throat.

Analytical and CFD analysis was confirmed with the measurements for both preliminary and final geometry in Turboinstitute’s laboratory.

Due to the very high system pressure (69 bar) structure analysis was part of the project. Mechanical design was made using ANSYS. The shape of the suction chamber is similar to an elbow and the consequence of the high pressure, were very high stresses at some critical areas. The final design of the suction chamber significantly depends on the mechanical design. Other parts of the jet pump are similar to the cylindrical shape and the stresses in the structure were not so high.

At the end, four full scale jet pumps were produced taking into account safety class 2.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 207

Heat transfer measurements in a single-phase flow of refrigerant R245fa in annular geometry

Boštjan Zajec, Marko Matkovič, Leon Cizelj

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

bostjan.zajec@ijs.si

 

As a part of the laboratory THELMA (Thermal Hydraulics experimental laboratory for multiphase applications) built at Reactor Engineering Division of Jožef Stefan Institute, a test section for studies of heat transfer in annular geometry was recently installed. The test section is designed to investigate various fluid flow and heat transfer regimes in annular geometry, ranging from single-phase to wide range of two-phase flow regimes. The first experiments will be performed with the refrigerant R245fa aiming to investigate the flow boiling under conditions resembling those in pressurized water reactor. The available experimental equipment enables visual observation of flow dynamics and measurements of local wall temperature profiles through the series of thermocouples integrated in the inner cylinder of the test section. The unique design of the test section consisting of the heating fluid (water) flowing through the inner cylinder and the primary fluid (refrigerant R245fa) enables rather accurate determination of local heat transfer along the test section.

In this study the experimental uncertainties related to heat losses in single-phase flow of refrigerant R245fa are quantified. Experimental measurements are performed over a wide range of flow and temperature conditions, also aiming to find an optimal configuration with the lowest overall experimental uncertainty. In each configuration systematic measurements of temperature profiles are performed. The experimental setup, calculated heat transfer coefficients and error analysis are presented.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 208

Comparison of 1D and 3D Water Hammer Analysis for Fast Transients in Nuclear Power Plant

Byung Soo Shin1, Gonghee Lee2

1Korea Institute of Nuclear Safety , 62 Gwahak-ro, Yuseong-gu, Daejeon, 34142, South Korea

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

bsshin@kins.re.kr

 

As mentioned in NUREG-0927(Rev.01), water hammer occurrence in nuclear power plants is one of unresolved nuclear safety issues. Water hammer can occur when the flow changes rapidly due to various unexpected causes. The generated pressure wave can lead to damage or malfunction of safety components by moving along the piping. As an example, HABIT 6, one of nuclear power plant in Korea, have experienced recently that pressure wave that is generated by unexpected malfunction of control valve in shutdown cooling system during cool down operation let LTOP(low temperature overpressure) relief valve being opened. One-dimensional water hammer analysis with simple and conservative assumptions about such fast transient showed that maximum peak pressure have been higher than the opening set pressure of the valve. In this study, it was evaluated that applicability of the typical method such as HANBIT 6 example used in one-dimensional analysis by the comparison of one-dimensional analysis and three-dimensional analysis. In particular, the difference of theoretical and experimental assumption about wave celerity and the application of experimental correlations for friction coefficient of flow obstructers such as orifice, control valve and etc. was evaluated in detail. One dimensional analysis was performed by FloMASTER, based on method of characteristics (MOC) and three dimensional analysis was done by FLUENT code, one of well-known commercial CFD code. Both results was compared with previous experimental results. Some meaningful findings was derived from analysis results. First, it was found that one dimensional analysis with experimental assumption for wave celerity predicted the experimental results well. However, three dimensional analysis had the almost same as the results from one-dimensional analysis with theoretical assumption for wave celerity in rigid pipe because of no assumption of elastic pipe. Second, experimental correlations of friction coefficient for simple circular pipe predicted experimental results well in one-dimensional water hammer analysis and also difference between one-dimensional analysis and three-dimensional analysis was meaningless if they have same flow conditions. But, if two orifices had different shapes in spite of same values of friction coefficient, it was found that the pressure wave characteristics could be different from each other during unsteady fast transient. Therefore, the unsteady friction coefficient based on inherent characteristics of each component should be considered in modelling for one-dimensional analysis.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 209

Flow and heat transfer CFD analysis in the test section of THELMA for wall surface temperature determination

Anil Kumar Basavaraj1, Marko Matkovič2, Blaž Mikuž2

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

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

blaz.mikuz@ijs.si

 

Thermal-Hydraulics Experimental Laboratory for Multiphase Applications (THELMA) has been built at Reactor Engineering Division of Jožef Stefan Institute. The main apparatus in the THELMA laboratory is a unique test section resembling a close proximity of a single fuel rod, which, among other things, allows visual observations of boiling phenomena including the Critical Heat Flux (CHF) in annulus geometry. Two fluids are involved in the test section: a primary fluid is the object of investigation (boiling, condensation, etc.) and a secondary fluid, which provides a controlled temperature of the rod’s wall. The flow and thermal conditions are monitored with flow meters and a number of miniature thermocouples embedded in the wall as well as installed within the secondary fluid path. However, despite they are numerous and miniature in size, they are still limited with the resolution as well as the insight into the flow and heat transfer behavior along the test section. Hence, Computational Fluid Dynamics (CFD) study is performed as a design support and for better understanding of flow and temperature distribution inside the test section. A series of CFD analysis is carried out for a wide range of single-phase test conditions, i.e. different flow rates, primary fluid temperatures, secondary fluid temperatures and heat transfer coefficients at the wall. Steady-state Reynolds Averaged Navier-Stokes (RANS) simulations are performed with a Conjugate Heat Transfer (CHT) approach in ANSYS Fluent CFD code. Obtained results are compared for low- and high-Reynolds turbulence models. The predicted temperatures at the surface of the fuel rod have been carefully examined and the uncertainty of such CFD prediction is estimated as well. Moreover, a simple correlation is derived from the obtained CFD predictions, which will be used as correction to the measured wall temperature for determining an accurate temperature at the surface of the rod.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 210

PIV measurements of turbulent flow over backward-facing step

Nejc Kosanič1, Boštjan Zajec2, Jure Oder2, Marko Matkovič2, Iztok Tiselj2

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

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

iztok.tiselj@ijs.si

 

Particle Image Velocimetry (PIV) is a technique for measuring liquid velocity fields. It is based on cross-correlation of subsequent images of microscopic particles that move with the liquid. An experiment with the geometry of backward facing step flow in a transparent test section with the length of around one meter is studied. Turbulent flows with Reynolds numbers around 7000 are studied in several horizontal and vertical cross-sections behind the step. Commercial system from LaVision with a single high speed camera and a pulse laser is used for a series of two-dimensional measurements of the velocity field at different cross-sections. The experimental setup and the results presented in this work are expected to be useful for comparison with accurate numerical simulations.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 211

Experimental observation of Taylor bubble disintegration in turbulent flow

Blaž Mikuž, Iztok Tiselj

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

blaz.mikuz@ijs.si

 

Two-phase flow exists in several regimes ranging from bubbly to annular flow. In a pipe, one of the regimes is slug flow, which consists of large intermittent gas structures separated by a liquid phase. These large intermittent gas structures are often referred to as Taylor bubbles, which have various shapes depending on the properties of the gas and liquid phase. Their behavior depends also on the background flow regime, i.e. laminar, transitional or turbulent, which has been extensively studied experimentally as well as numerically.
Recent large-eddy simulations of a Taylor bubble in a turbulent flow regime revealed an interesting and very complex behavior of the flow in the wake of a Taylor bubble. In that region, the gas-liquid interface exhibits violent flapping. Moreover, from the tip of the gas-liquid interface small bubbles are generated, which get trapped in a recirculation zone. Some of them are merged and joined back with the main bubble whereas the others are transported downstream the flow and lost. For that reason, the Taylor bubble at some point completely disintegrates. The rate at which that happens depends strongly on the correct prediction of the small bubbles production, i.e. tearing from the main bubble, which is quite challenging for numerical simulations. Therefore, in the present study this phenomenon is experimentally investigated in a test facility, where Taylor bubble is exposed to a counter-current flow that equalizes the buoyancy of the bubble. The mixture consists of liquid water and air, which allow visual observations through the glass pipe walls. Such data are needed for better understanding of the phenomenon itself as well as for a validation of the high-fidelity simulations.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 212

Investigations on flow boiling under forced convection on zircaloy tubes up to critical heat flux

Walter Tromm

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

walter.tromm@kit.edu

 

Boiling under forced convection is an important mechanism for heat transfer in nuclear power plants that can take place under operating conditions as well as in various accident scenarios. The heat transfer is limited by the critical heat flux (CHF) which can be measured when the boiling crisis occurs.
Required is a predictable, mechanistic simulation code which is able to calculate the boiling behavior of a system up to the CHF independent from the boundary conditions (mass flow, liquid temperature, pressure, geometry …). The development of a predictive code requires detailed validation measurements under different conditions and different geometries to describe the behavior of the flow and the heat transfer precisely. This contribution presents the results of an extensive measurement campaign. The low pressure facility COSMOS-L and the experimental program will be presented. The facility operates with subcooled water as working media at two different test geometries, a single heated rod in the annulus and a rod bundle. System pressure, subcooling and massflow were varied as well as the heating power to get a sufficient dataset.
In 55 CHF experiments, a statistical CHF study was carried out under constant boundary conditions. The resulting frequency distribution shows the scattering of the CHF-values, which are presumably caused by turbulence and phase interactions, for example during the bubble detachment at the heater tube. A following high speed study gives a visual impression of the flow under different flow conditions. The basic CHF-data is then supplemented by temperature distributions and boiling curves. In addition, velocity profiles and bubble characterization near the heated surface were measured with by (laser-) optical measurement methods like laser doppler anemometry or laser-assisted shadowgraphy.
In addition to the results the presentation gives a short outlook to the upcoming high pressure experiments on the new test facility COSMOS-H. COSMOS-H is a high pressure facility and achieves prototypical thermohydraulic conditions (17 MPa, 360°C). With a system power of 2 MW the facility offers the opportunity to investigate flow boiling up to CHF on single tubes and rod bundles with a maximum test section power of 600 kW. A heated length of 4 m is possible.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Thermal-hydraulics – 213

BEYOND DESIGN BASIS ACCIDENTS WITH INTERLOOP LEAKAGE INDUSED BY THE HYDRAULIC SHOCK DAMAGING THE VVER STEAM GENERATOR

Volodymir Skalozubov, Victor Kolykhanov, Igor Kozlov, Denis Pierkovsky, Oleg A. Chulkin, Yuri Komarov, Irina Aretinskaya

Odessa Polytechnic University, Shevchenko av. 1, 65044 Odessa, Ukraine

victor.kolykhan@i.ua

 

One of the reasons for the multiple equipment failures in case of beyond design basis accidents may be the occurrence of hydraulic shocks (HS), accompanied by pulsed high-amplitude hydrodynamic loads on equipment and elements of the NPP piping systems. The fundamentally important effect of the flow kinetic energy transition into HS during deceleration do not taken into account in known models.
Based on computational studies the critical conditions for the HS occurrence during accelerated closure of valves during beyond design basis accident have been determined. Besides the conditions for the occurrence of HS were determined under design-basis and beyond design-basis accidents scenarios with interloop leakage for varying leakage size. The effect of shut-off valve on the critical conditions for HS occurrence in the two-phase flow mode is analyzed.
The performed computational simulation showed that the state of the emergency feed water system (EFWS) for steam generator is sufficiently close to the boundary of critical conditions for HS occurrence. Therefore, as a compensating measure to improve the reliability of the system, to install additional throttle devices on the injection section of the main line of EFWS is proposed.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Materials, Integrity and Plant Life Management – 304

Neutron-irradiated nanocrystalline silicon carbide (3C-SiC) particles investigation by EDP method

Elchin M. Huseynov

Institute of Radiation Problems National Academy of Science of Azerbaijan, F. Aghayev 9, AZ 1143 Baku, Azerbaijan

elchin.h@yahoo.com

 

Over the past few years, obtaining new materials or reprocessing current materials in the nuclear and cosmic technologies have been at the focus of various studies worldwide. SiC is an attractive material that can be applied in nuclear technologies due to the physical and chemical properties [1-5]. High-temperature resistance, high-perfection structure, mechanical stability, high oxidation resistance increase the application potential of SiC as a nuclear material [2-4]. The combination of perfect mechanical and functional properties is the basis of an application of SiC as a semiconductor in modern electronics.
At the present work nanocrystalline silicon carbide (3C-SiC) has been irradiated with neutron flux (~2x1013n•cm-2s-1) up to 20 hours at different periods. Electron diffraction patterns (EDP) investigation of nanocrystalline 3C-SiC particles (~18nm) is comparatively analyzed before and after neutron irradiation. The effect of irradiation on the crystal structure of the nanomaterial was studied by selected area electron diffraction (SAED) and EDP analysis. Amorphous transformation effects of neutron irradiation on the nanocrystalline silicon carbide (3C-SiC) have been studied by EDP method.
It is clear from EDP analysis that nanocrystalline 3C-SiC particles have little amorphous transformation after neutron irradiation. Probably, after neutron irradiation, there is a thicker amorphous layer surrounding the 3C-SiC nanoparticles. The amorphous layer of the surfaces of nanoparticles can cause a greater or lesser agglomeration degree. As a result of EDP analysis it has been found out that, by a majority, neutron irradiation doesn’t affect the crystal structure of 3C-SiC nanoparticles (excluding small amorphous transformation).

1. I.Vivaldo, M.Moreno, A.Torres et al. “A comparative study of amorphous silicon carbide and silicon rich oxide for light emission applications” Journal of Luminescence 190, 215-220, 2017
2. YutaiKatoh, Lance L. Snead, IzabelaSzlufarska, William J. Weber “Radiation effects in SiC for nuclear structural applications” Current Opinion in Solid State and Materials Science 16, 3, 2012, 143–152
3. Elchin Huseynov, Anze Jazbec “Trace elements study of high purity nanocrystalline silicon carbide (3C-SiC) using k0-INAA method” Physica B: Condensed Matter 517, 30–34, 2017
4. MohdIdzatIdris, Hiroshi Konishi, Masamitsu Imai et. al “Neutron Irradiation Swelling of SiC and SiCf/SiC for Advanced Nuclear Applications” Energy Procedia 71, 2015, 328–336
5. Elchin M. Huseynov “Investigation of the agglomeration and amorphous transformation effects of neutron irradiation on the nanocrystalline silicon carbide (3C-SiC) using TEM and SEM methods” Physica B: Condensed Matter 510, 99–103, 2017






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Materials, Integrity and Plant Life Management – 305

Statistical study of intergranular stresses in untextured polycrystalline metals

Samir El Shawish

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

samir.elshawish@ijs.si

 

Mechanical stresses between the crystalline grains may induce damage by intergranular cracking in a metallic polycrystalline aggregate. Stress amplitudes on grain boundaries depend not only on the external loading but also on the material properties of the grains, grain shapes and crystallographic grain orientations. For a particular metallic aggregate, these stresses can be accurately estimated using computationally demanding simulations (based on, e.g., crystal plasticity finite element method), however, such an approach deems impractical for a general case. In a recent study, a new empirical relation – a shortcut – has been identified to quickly and accurately estimate the scatter of intergranular normal stresses present in a general random polycrystalline metal when exposed to external mechanical loading (e.g., tensile or biaxial). Using this scatter relation, the probability for crack initiation may be reliably obtained for a given grain boundary strength, constituting a tool for classifying polycrystalline random aggregates according to their potential susceptibility to intergranular stress corrosion cracking.

In this work, the study is extended to account also for polycrystalline metals with non-random crystallographic grain orientations. It is well known that different types of grain boundaries do show different cracking sensitivities: e.g., random high-angle grain boundaries seem to be more prone to cracking than low-angle grain boundaries (e.g., ?3 in the Coincidence Site Lattice theory). In this view, a possible influence of grain boundary type on intergranular normal stresses is investigated and discussed.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Materials, Integrity and Plant Life Management – 306

Modeling cladding oxidation with coupled thermal-mechanics and thermal-hydraulics solvers

Henri Loukusa, Jussi Peltonen

VTT Technical Research Centre of Finland Ltd., Kivimiehentie 3, 02044 VTT, Finland

henri.loukusa@vtt.fi

 

The zirconium cladding in typical light-water reactor nuclear fuels oxidizes during normal operation forming a protective oxide layer. The oxide layer has a lower thermal conductivity than that of the metal, and therefore affects heat transfer from within the fuel rod. The temperature of the cladding is most important in determining the extent of cladding oxidation. However, the cladding temperature is not solely determined by the thermal behavior of the fuel rod itself, but in large part by the heat transfer conditions from the cladding to the coolant. This coupled phenomenon provides a suitable case for validating the predictions of a coupled simulation including fuel rod thermal mechanics and coolant thermal hydraulics.

The fuel behavior module FINIX has been developed at VTT for providing a reasonably accurate description of fuel thermal-mechanical behavior in coupled applications. In addition, quite recently a new thermal-hydraulics solver Kharon has been developed. The closed channel two-phase steady state solver is based on the porous medium approach and in this work it provides the description of cladding-coolant heat transfer. From thermal hydraulics the steady-state axial temperature distribution in the cladding is obtained, which then determines the axial variation in the cladding oxide thickness in the fuel behavior description.

A pressurized water reactor assembly from the literature is modeled with both codes and the resulting cladding oxidation predictions are compared to experimental data. A somewhat idealized linear heat generation rate history is used for the rods in the assembly, as Kharon models the system at the assembly level and FINIX at the rod level. The predictions of this approach are compared to the oxidation predictions of standalone FINIX, which contains basic thermal-hydraulic correlations for the calculation of the axial temperature distribution in the cladding.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Materials, Integrity and Plant Life Management – 307

Prototyping of a harsh conditions resistant ultrasound NDE tool for water level measurements in a nuclear power plant primary circuit

Nikola Bunčić, Marko Budimir

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

nikola.buncic@inetec.hr

 

It was shown several decades ago that many nuclear power plants were inadequately equipped for complete operations control in the outage conditions that comprise temporarily dissimilar configurations and system parameters. This made the predicted general risk of an accident to be the highest exactly during the outage. Within the maintenance service period the level of water in the plant primary circuit can be lowered down below the levels covered by standard instrumentation, originally designed for the control of the circuit while it is in regular operation conditions. The most critical is the state of the pumps for remnant heat dissipation out of the nuclear reactor when the water level within the primary circuit pipeline is lowered (mid-loop operation). For the water level monitoring purposes, an additional advanced NDE system thus needs to be installed. The currently used ultrasound monitoring tools, mounted directly to the hot leg pipe coming out of the reactor vessel, either have problems with the measurements procedure stability, or they need to be removed after every outage and installed at the beginning of every next outage, which complicates the service procedure and creates personnel health risks (additional scaffolding needed, complicated access location, work at heights, additional ionizing radiation doses, …).
This work presents a prototyping procedure, and its experimental results, of a mechanically active long term operating ultrasound NDE tool for measurements of water levels in the hot leg pipe of the primary circuit of a nuclear power plant during the outage procedure. The system functionality is based on a high temperature resistant piezoelectric-based air-coupled ultrasound transducer and an active mechanical ring holding the transducer and having states in which the transducer is directly attached to the pipe (during outage) and in which the transducer is away from the pipe (during plant normal operating state). The system is designed not to be removed and reinstalled from the hot leg pipe once it is initially installed and to be used for at least several outages in a row without a need to access the transducer on the pipe. A complete virtual prototype (CAD and FEM) model of the system design and its electromechanical behavior is presented, together with laboratory experiments on the transducer and its materials properties.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Severe Accidents – 406

Analysis of Severe Accident in Safety Upgraded Krško NPP with MELCOR 2.2

Matjaž Leskovar, Mitja Uršič

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

matjaz.leskovar@ijs.si

 

Following the lessons learned from the accident at the nuclear power plant Fukushima in Japan and according to the Slovenian Nuclear Safety Administration decree, the Krško NPP decided to take the necessary steps for upgrading the safety measures to prevent severe accidents and to improve the means for the successful mitigation of their consequences. One of the modifications that the Krško NPP will implement is the installation of an alternative safety injection pump and an alternative residual heat removal heat exchanger. This modification will, among the other already existing systems, serve for the purpose of reactor decay heat removal, either from the reactor coolant system (RCS) or from the containment, once the core and RCS are severely damaged.

The purpose of the paper is to analyse the response of the Krško NPP containment following a severe accident taking into account mitigation measures for heat removal from the containment solely by the planned alternative safety systems. As the initiating event, a strong earthquake was considered, resulting in a simultaneous station black out and large break loss of coolant accident. Three scenarios were analysed: (1) no mitigation, (2) alternative safety systems available 24 h after initiating event with water injection through containment sprays, (3) alternative safety systems available 24 h after initiating event with water injection into RCS. The analyses were performed with the MELCOR 2.2 computer code.

In the paper the simulation results will be presented and thoroughly discussed in comparison to the previous study performed with the MELCOR 1.8.6 code and the study performed with the MAAP 4.0.7 code. The results of the performed calculations with MELCOR 2.2 show, in opposite to the MELCOR 1.8.6 calculations, that by flooding the reactor cavity it is possible to stabilize the poured molten core, but only if the molten core concrete interaction has not started yet. The main results of the MELCOR 2.2 and MELCOR 1.8.6 calculations are similar, but significantly differ from the MAAP 4.0.7 calculation results, and consequently the conclusions regarding the most appropriate severe accident management are different. The differences in the modelling approach resulting in the different results of the codes will be explained.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Severe Accidents – 407

Simulation of premixed layer formation in stratified melt-coolant configuration

Janez Kokalj

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 has the potential for causing severe core damage, including core meltdown. If the hot melt comes in contact with the coolant water, the internal energy is rapidly transferred, which can result in a steam explosion. Considering the amount of thermal energy, initially stored in the liquid corium melt at about 3000 K, this phenomenon can jeopardize the integrity of the containment and can cause damage to the systems inside. Consequently, the possibility of a radioactive leakage presents danger for the environment and general public safety. Similar explosion phenomena can be a concern in some industrial processes, such as foundries and liquefied natural gas operations or in certain volcanic activity where water is present.
In severe accident analysis in nuclear power plants, when studying fuel-coolant interactions, a melt jet pouring into a coolant pool is the geometry most widely studied. In the complementary stratified configuration, a continuous layer of melt lies beneath water and both layers are separated by a steam film. This configuration was believed to be incapable of creating a significant premixed layer and producing strong explosions, based on some experimental and analytical work from the past. However, the results from recent experiments performed at the PULiMS and SES facilities (KTH, Sweden) with corium simulants materials contradict this hypothesis. A clearly visible premixed layer and strong spontaneous vapour explosions were observed in some of the tests.
In the paper, simulation of the premixed layer formation in the stratified melt-coolant configuration will be discussed. A modelling approach for the premixed layer formation was developed and incorporated in the thermodynamic code MC3D. The MC3D computational code is devoted mainly to the simulations of fuel-coolant phenomena and this new application of the code will be presented. Numerical tests, applying the developed model, were performed and the results will be analysed and discussed.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Severe Accidents – 408

Simulations of heat and mass transfer around circular core fragment in sodium coolant

Matej Tekavčič, Mitja Uršič, Matjaž Leskovar

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

matej.tekavcic@ijs.si

 

During a hypothetical core melt accident in innovative sodium cooled fast reactors, the rapid and intense heat transfer interaction between the molten core material and the sodium coolant may lead to a vapour explosion. The precise nature of the pressurization process amid the explosion in sodium is unknown. The conditions during the explosion phase are hard for an adequate experimental investigation of the heat and mass transfer process. Another difficulty is the opaqueness of sodium. Thus, analytical research and precise numerical simulations could support the investigation of the pressurization process in sodium.

The present work focuses on two-dimensional simulations of heat and mass transfer around a circular fragment of melted core in liquid sodium coolant. The paper considers the forced convection film-boiling regime assuming a large interface between the vapour and liquid sodium phases for the heat and mass transfer model. The initial and boundary conditions represent the expected flow conditions during a vapour explosion in sodium: operating pressures up to 10 MPa, fragment velocities up to 20 m/s and fragment size of 0.1 mm. The results include sensitivity studies on mesh configuration, turbulence model and heat and mass transfer model. Our final objective is to assess the applicability of the Epstein-Hauser heat transfer correlation for modelling of heat and mass transfer processes in sodium.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Severe Accidents – 409

Simulation of Hydrogen Combustion Experiment in THAI+ Facility with ASTEC Code

Ivo Kljenak, Rok Krpan

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

ivo.kljenak@ijs.si

 

Hydrogen combustion is one of the phenomena which may occur during a severe accident in a light water reactor nuclear power plant, as hydrogen generated due to oxydation of the uncovered reactor core would flow into the containment and its concentration in some parts could exceed the flammability limit. This phenomenon is being studied both experimentally and theoretically. The THAI+ experimental facility (located at Becker Technologies, Eschborn, Germany), consists of a larger cylindical vessel and a smaller paralled attached drum, with the two vessels linked both at their upper and lower ends with wide connecting pipes. Experiments on hydrogen combustion, performed in this facility, provide new insights into possible flame propagation through the compartments of an actual nuclear power plant containment. Although lumped-parameter codes cannot simulate flame propagation with the same accuracy as Computational Fluid Dynamics (CFD) codes, their use is still valuable, as they might be used for simulation of combustion in actual containments, where the use of CFD codes is impractical due to the large numerical meshes (in terms of number of cells). For this reason, the validation of lumped-parameter codes using experimental results of hydrogen combustion is beneficial for safety analyses of NPPs. One such code is ASTEC, which is being developed by the Institut de Radioprotection et de Surete Nucleaire (France), where the module that simulates containment phenomena uses a lumped-parameter description.

In the proposed paper, the simulation of the experiment on hydrogen combustion THAI HD-36 with the ASTEC code is presented. In the experiment, a mixture of air, hydrogen and steam was ignited at the bottom of the main vessel. The flame then travelled both upwards through the vessel and to the attached drum through the lower connecting pipe. A multi-volume model of the THAI+ facility for the ASTEC code was developed. The simulated pressure, temperatures and flame propagation are compared to experimental results and discussed.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

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

 

M. Kunšek (JSI), I. Kljenak (JSI), L. Cizelj (JSI)
During a hypothetical severe accident in a light water nuclear power plant, the reactor 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 pool scrubbing phenomena.
In the proposed paper, the behavior of transition of solid particles from gas phase to liquid phase during the bubble rise in a scrubbing pool was simulated using subgrid modeling. The multi-phase simulations were performed for particles, bubbles and liquid using the open-source Computational Fluid Dynamics code OpenFoam, with the solver reactingMultiphaseEulerFoam. In the simulation, the gaseous, liquid and two particle phases (phase 1 within bubbles and phase 2 within liquid) were simulated. All phases were described in 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 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 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.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Severe Accidents – 411

MELCOR Simulations of the SBO in Gen III PWR with EVMR

Piotr Darnowski1, Piotr Mazgaj1, Eleonora Klara Skrzypek2

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

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

piotr.darnowski@pw.edu.pl

 

The paper presents a study of the Station Blackout accident in a generic 4-loop large PWR reactor with Ex-Vessel Melt Retention (EVMR). The nuclear power plant is equipped with core melt stabilisation device (core-catcher) which plays the primary role in its EVMR strategy. The plant model and simulations were prepared with the MELCOR 2.2 computer code for both in-vessel and ex-vessel phases. The new MELCOR Lower Head Containment package was applied to simulate the core catcher device. Parametric type sensitivity calculations were performed to study system response to different core catcher assumptions. The emphasis was put on the containment long term response. Finally, conclusions are found about the EVMR strategy and core catcher modelling with MELCOR.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Severe Accidents – 412

Reactor Vessel Modelling with the MELCOR Code

Siniša Šadek, Davor Grgić, Franjo Vuković, Vesna Benčik

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

sinisa.sadek@fer.hr

 

MELCOR is an integral severe accident code that enables calculation of complete transient sequence in a reactor coolant system (RCS), nuclear power plant (NPP) secondary side and the containment. NPP modelling with the MELCOR code provides the user freedom to develop its own modelling approach by following general guidelines, or rules, determined by code developers. Thus, the nodalization is not prescribed but it is still important to create a model in a logical manner that reflects the system design and operation.
The code was used for a calculation of a postulated station blackout accident in the NPP Krško. The whole sequence of events during a severe accident was covered, including the thermal-hydraulic behaviour of the RCS, core cladding oxidation, fuel elements degradation and melt-down, molten corium concrete interaction in the containment cavity, the containment heat-up and pressurization. The focus of the paper was on the reactor core behaviour and influence on timing of core damage propagation events caused by different reactor pressure vessel (RPV) and core thermal-hydraulic models. Two models of the RPV were developed: one that uses complex nodalization of the RPV upper plenum, the core and the lower plenum, and the other one that uses a coarse nodalization scheme for these components. For example, in the first NPP model, the core was modelled with 12 axial control volumes, while in the second one, it was modelled with only one control volume representing the whole core region. In both models, the core components were radially divided in five regions. The similarities and differences in the analysis results between those two approaches will be presented and discussed in the paper, with special attention given to numerical aspects of the calculation, like the time-step selection and the CPU time requirement.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Severe Accidents – 413

NPP Krško Large Break Loss of Coolant Accident Calculation using MELCOR 1.8.6 and MELCOR 2.2 Codes

Vesna Benčik, Davor Grgić, Siniša Šadek, Štefica Vlahović

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 1.8.6 is 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 new code version MELCOR 2.2 and both input decks are currently being tested and verified. In this paper the results of Large Break Loss of Coolant Accident (LB LOCA) using both MELCOR 1.8.6 and MELCOR 2.2 are presented. Both unmitigated scenario (engineered safety features not available) and mitigated scenario (one train Emergency Core Cooling System 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%). Analysis of LB LOCA accident for both MELCOR 1.8.6 and MELCOR 2.2 has demonstrated satisfactory behavior of ESFs (1 ECCS train and 1 ESF train in containment), both in RCS and in the containment. The results obtained with both MELCOR versions show similar qualitative and quantitative behavior.

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






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Research Reactors – 505

Application of JSIR2S for dosimetry calibration of Nürfet semiconductor dosimeters at the JSI TRIGA reactor inbetween reactor shutdowns.

Klemen Ambrožič1, Luka Snoj1, Gregor Kramberger2

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

2Jožef Stefan Institute, Experimental Particle Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

klemen.ambrozic@ijs.si

 

The JSIR2S code for calculations of delayed radiation fields has been in the validation and testing stages of development for the last year, where Its initial validation on a simple ITER port plug computational model has been presented at the NENE conference. As the analog MCNP coupled neutron-photon accounts only for prompt gamma-rays in steady state operation, the JSIR2S code is used for delayed gamma field calculations, accounting for remaining 20-30% of gamma rays, which remain even after shutdown. This time we present how the code can support irradiations carried out at the JSI TRIGA reactor by combining modelling and measurements for dose calculations received by irradiated samples. Each time ionization chamber measurements were performed to evaluate the accuracy of calculation results.
Semiconductor Nürfet dosimeters were irradiated at least one hour after reactor shutdown in order for the delayed neutron precursors to decay leaving predominantly gamma field radiation. In this paper we present measurement and calculation data to support Nürfet response evaluation and initial stages of their calibration in representative reactor gamma spectra.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Research Reactors – 506

Jožef Stefan Institute TRIGA Research Reactor Activities in the Period from September 2018 – August 2019

Anže Jazbec1, Sebastjan Rupnik1, Andraž Verdir1, Marko Rosman1, Vladimir Radulović2, Luka Snoj2, Borut Smodiš3

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

 

Jožef Stefan Institute (JSI) operates a 250 kW TRIGA research reactor for more than 53 years. Safety performance indicators (SPI) are monitored for over 10 years now. Some of the monitored SPIs are operating time, number of irradiated samples, doses received by operating staff and activity of noble gases released to the environment. In the paper, SPIs for the year 2018 are presented and compared to the ones from previous years. Upon the SPIs analysis, future operation can be improved and safety of the reactor can be increased.
Furthermore, new research work carried out during the years 2018 and 2019 is presented. Two research campaigns for CEA (French Alternative Energies and Atomic Energy Commission) were performed. In the first one, activation of various materials with fast neutrons was determined. In the second one, CEA staff was investigating self-powered neutron detector responses. Similar campaign was done for the French company THERMOCOAX. In collaboration with three UK research companies, response of ultrasonic detectors to neutron and gamma radiation was determined. In June 2019, irradiation campaign for a NATO-funded project is planned and the results will be discussed in the presentation.
Finally, modifications done at the reactor during the last year are presented. The most extensive one was pneumatic transfer system digitalization. The existing pneumatic system of the carousel neutron irradiation facility was connected with the standard pneumatic transport system in the irradiation channel F24. Next modification, completed in the year 2019, was digitalization of control rods positioning. The number, assigned to a particular rod position, is now shown on the control board and stored in the reactor’s digital archive. The last important modification was a new triangular irradiation channel. It is located in the same place as the old one but it is significantly bigger.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Research Reactors – 507

Evaluation of Uncertainties in Theoretical Predictions of Pulse Mode Operation

Ingrid Vavtar, Luka Snoj

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

ingrid.vavtar@ijs.si

 

The TRIGA reactor at the »Jožef Stefan« Institute can operate in stationary and in pulse mode. In pulse mode a transient control rod is quickly ejected out of the reactor core by pneumatic mechanism. After ejecting reactor becomes prompt supercritical in a short time, a few tens of ms and the power begins to increase exponentially. When the power and consequently the fuel temperature increase the reactivity decrease due to the prompt negative temperature reactivity coefficient of the fuel. A decrease of the reactivity results in slowing down the chain reaction and consequently in a decrease in power, which makes the reactor establish in a new equilibrium state quickly. The peak power of the pulse reaches some 100 MW and the total relaxed pulse energy is relatively small (typically a few MJ) due to the short pulse time. The pulse mode is mostly used to validate the computational models, for education purposes and for demonstration of inherent safety prompt negative temperature reactivity coefficient of the reactor. Moreover, pulse mode operation is also used to perform radiation hardness tests of electronic components.
The data from all pulse experiments that have been carried out on the TRIGA Mark II reactor at the »Jožef Stefan« Institute have already been collected and are available at http://trigapulse.ijs.si/. The purpose was to analyse all the pulses that have been performed so far and to check the matches between measurements and theoretical predictions. Theoretical predictions are from the Fuchs-Hansen (FH) model and Nordheim-Fuchs (NF) model, which have the same limit values, maximal power, total released energy and full width at half maximum. In order to compare calculated and measured physical parameters it is important to estimate their experimental and computational uncertainties. The FH model was used for evaluation of uncertainties in reactor physical parameters (peak pulse power, FWHM, total released energy) due to uncertainties in physical quantities, such as inserted reactivity, effective delayed neutron fraction, average lifetime of prompt neutrons and effective temperature reactivity coefficient of fuel. In addition correlation of physical parameters in theoretical predictions was analysed. The highest correlation of 90 % was estimated between effective delayed neutron fraction and average lifetime of prompt neutrons obtained from the same experiment. All correlations were estimated conservatively in order to assess the greatest differences between correlated and uncorrelated uncertainties. It was found that despite of the large estimated correlations of parameters, the correlations do not contribute significantly because of multiplying contributions of different sizes. Therefore for the analysis it is sufficient to take into account that the parameters are uncorrelated.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 606

Impact of Burnable Absorbers on Nuclear Data Uncertainty Analysis for Fuel Assembly Depletion

Martin Lovecky, Jiri Zavorka, Jana Jirickova, Radek Škoda

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

lovecky@rice.zcu.cz

 

Nuclear data uncertainty propagation can be analyzed for various applications. Uncertainty of neutron multiplication factor during fuel depletion is the response quantity of interest for burnable absorber research. Three different fuel assembly designs were analyzed, FA without burnable absorber, FA with gadolinium bearing rods in selected positions and finally, FA with uniform BA loading in all fuel pins. Newly developed BA materials would be introduced faster into reactor operation if they improve not only neutronics, but other safety or economics-related issues, i.e. accident tolerant fuel. Uniform BA loading is required for technical-economical analyses. VVER-440 fuel assembly depletion with non-uniform gadolinium and uniform erbium BA designs were analyzed with SCALE code package in order to show the effect of BA design on reactivity uncertainty. Nuclear data uncertainty analysis methods can be divided into first order perturbation theory based methods (TSUNAMI in SCALE) and Monte Carlo sampling methods (SAMPLER in SCALE). The first method has the advantage in fast calculation time due to the use of adjoint-based sensitivity approach, on the other hand, it is efficient only for a sufficiently small number of responses. The latter method requires performing large number of costly transport calculations, but can become quite competitive for complex models.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 607

Study on probability distribution of input nuclear data in random sampling method for uncertainty quantification calculation of reactor physics parameters

Ryotaro Kimura

Hokkaido University of Education, 1-2 Hachiman-cho, Hakodate 040, Japan

jonhamuro@eis.hokudai.ac.jp

 

1.Background and purpose of this study
The random sampling method is one of methods to evaluate uncertainty of nuclides composition after nuclear fuel burnup induced by uncertainty of nuclear data. Only variance and covariance data are defined in evaluated nuclear data files, and the probability distribution followed by the nuclear data is not specified. Therefore, in general, calculations are performed assuming a normal distribution. However, some nuclear data give a standard deviation of more than 50%, and in such a case, there is a problem that negative values are sampled in random sampling. In order to cope with this problem, in this study, we calculate fuel burnup assuming lognormal distribution to nuclear data and evaluate the influence on the evaluation result.

2. Evaluation of uncertainty in random sampling method
In this study, statistics of output parameters are calculated by random sampling method. We consider higher order moments such as skewness andsup< of the probability distribution as an indicator of the effect of differences in probability distribution.
The lognormal distribution is named because the logarithm of a random variable according to this distribution follows the normal distribution, and it does not have negative values.

3. Calculation method
Uncertainty information such as mean value and standard deviation is acquired from nuclear data files, and a covariance matrix is created using them. The sampling is performed based on the determined covariance matrix, and a fuel burnup chain is created. Finally, fuel burnup chain is performed using the above data to obtain nuclide number density data, and the statistics of number density after fuel burnup are evaluated by statistical processing. In this study, fuel burnup calculations are performed assuming that a light water reactor loaded with uranium oxide fuel is an infinite array single cell model.

4. Calculation result, consideration
Assuming a lognormal distribution for half-life, the skewness of some nuclides is inclined in the negative direction compared to the case assuming normal distribution. The common point of these nuclei is that the relative standard deviation of the number density is large.
Assuming a lognormal distribution for fission yield without correlation, the effect of uncertainty is that the standard deviation becomes smaller, and some nuclides have been skewed in the positive direction of skewness and kurtosis.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 608

Pin power reconstruction method for rectangular geometry in nodal neutronics program Ants

Antti Rintala1, Ville Sahlberg2

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

2VTT Technical Research Centre of Finland Ltd., Kivimiehentie 3, 02044 VTT, Finland

antti.rintala@vtt.fi

 

VTT Technical Research Centre of Finland Ltd (VTT) is developing a new computational framework, called Kraken, for reactor core multi-physics problems. The framework consists of modular neutronics, thermal hydraulics and thermal mechanics solvers coupled together via a central multi-physics driver module. Ants is a novel reduced order nodal neutronics program developed as a part of Kraken. The published methodology and first results of Ants have previously been limited to steady state multigroup nodal diffusion solutions in both rectangular and hexagonal geometries.

Ants solves the multigroup nodal neutronics using a hybrid of analytic function expansion nodal (AFEN) and flux expansion nodal (FENM) methods using a set of analytic basis functions for each mode flux. The pin powers are reconstructed by modulating the homogeneous pin cell fluxes obtained with direct integration of the homogeneous nodal fluxes over the pin cells with the heterogeneous group-dependent pin power form functions. The form functions are obtained from single assembly calculations using the continuous-energy Monte Carlo particle transport code Serpent 2.

This work describes the Ants pin power reconstruction methodology for rectangular geometry. Initial comparisons with the heterogeneous solutions of Serpent are presented in simple configurations.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 609

Gamma and Neutron Dose Rate Calculation Around the Jožef Stefan TRIGA Research Reactor using hybrid methods

Anže Jazbec1, Bor Kos2, Luka Snoj2

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

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

anze.jazbec@ijs.si

 

Knowledge on dose fields around nuclear facilities is important to ensure a safe working environment for operating staff, researchers and also visitors. In case of Jožef Stefan Institute TRIGA research reactor, staff responsible for radiation protection does regular weekly survey of the radiologically controlled area. Survey consists of contamination check and dose rate measurements. However, in some cases it is not possible to perform direct measurements.
For example, in case of planning a new experiment or new radiation shield it is not possible to do the measurements, since the equipment is not installed yet. In that case, it is crucial to perform some scoping calculation in order to show that proposed shielding is sufficient. Similar goes for accident scenarios, for example loss of coolant accident. Because of safety concerns the pool cannot be simply drained to measure dose rates. Therefore, it is important to have a reliable tool that can be used to calculate neutron and gamma dose rates.
A computation procedure based on the Monte Carlo N-Particle (MCNP) Transport Code accelerated using the ADVANTG code has been implemented. The whole JSI TRIGA reactor hall was modelled including walls, roof, basement, spent fuel pool and control room. Irradiated fuel was taken as a source of radiation – according to the results, this is sufficient since the fuel is the major contributor of the source neutrons and gamma rays. Prompt neutron, prompt gamma and delayed gamma can be simulated. The method was already verified on two cases in order to validate prompt and delayed source. As an example of usage, LOCA was simulated and gamma dose rates were calculated for the reactor platform and control room at several times after reactor shutdown.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 610

The Results of Studies on Water-Moderated Lattices and Program of Experiments on Fast Multiplying Systems with 19.75% Enriched UZrCN Fuel

Svyatoslav Sikorin1, Siarhei Mandzik1, Siarhei Polazau1, Andrei Kuzmin1, Tatsiana Hryharovich1, Yury Damarad1, Iryna Soltan1, Shamil Tukhvatulin2, Igor Bolshinsky3, Yousry Gohar4

1Joint Institute for Power and Nuclear Research – Sosny of the National Academy of Science of Belarus, 47/22 Prilesye district, Lugovaya Sloboda village council, Minsk district, 223063 Minsk region, Belarus

2The Joint Institute of Power and Nuclear Research – Sosny of the National Academy of Sciences of Belarus , PO BOX 119, 220109 Minsk, Belarus

3Scientific and Industrial Association “RADON”, 7th Rostovsky Lane, P.O.Box 2/14, MOSCOW 119121, Russian Federation

4Idaho National Engineering and Environmental Laboratory, 2525 North Fremont Ave., P.O.Box 1625, Idaho Falls, ID 83415, USA-Idaho

5Argonne National Laboratory, Building 223, 9700 South Cass Avenue, ARGONNE ILLINOIS 60439, USA-Illinois

sikorin@sosny.bas-net.by

 

UZrCN fuel is a high density, high temperature fuel that has potential for application in different types of reactors, comprising research reactors, including of conversion HEU on LEU fuel. Criticality experiments on U-H2O lattices with LEU UZrCN fuel rods were performed using “Giacint” critical facility of the Joint Institute for Power and Nuclear Research –Sosny of the National Academy of Science of Belarus. Five configurations of critical assemblies were studied. The fuel rod pitch is 18, 32 or 47 mm with triangular grid. The criticality conditions were obtained by adjusting the water moderator height in the studied lattices. The fuel was designed and produced by the Scientific Research Institute Scientific Industrial Association “LUCH” under the Russian Research Reactor Fuel Return Program of the U.S. Department of Energy. The fuel material UZrCN has U-235 enrichment of 19.75%. The uranium metal density is 10,8 g/cm3. The total fuel rod length is 620 mm, and the active fuel length is 500 mm. The outer fuel rod diameter is 12 mm , and the fuel material diameter is 10.7 mm. The clad material is stainless steel or niobium. The results of experiments on critical assemblies have been analyzed by creating detailed calculation models and performing simulations for the experiments. This paper presents the obtained experimental and analytical results.
The “Giacint” and “Kristal” Critical facilities will be used to determine neutron-physical characteristics of critical and subcritical assemblies on the fast neutron , modeling physical features of cores of advanced reactor and accelerator driven systems, cooled gas, and liquid-metal coolants. The critical and subcritical assemblies represent uniform hexagonal lattices of fuel assemblies ( 39 mm pitch), each of which consists of 7 fuel rods and have no cladding. Three types of fuel assemblies with different matrix material (air, aluminum and lead) were investigated. The side radial reflector is Be (internal layer) and stainless steel (external layer). The description of the design and the composition of the critical and subcritical assemblies with UZrCN fuel, the results of calculation are presented.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 611

Nuclear data adjustment based on Bayesian inference applied to covariance matrix generation

Florian Batard1, Ivo Aleksander Kodeli2, Pierre-Jacques Dossantos-Uzarralde3

1Multiple organizations possible, Unknown, Unknown, Slovenia

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

3École Nationale Supérieure d’Informatique pour l’Industrie et l’entreprise, 1, Square de la Résistance, F-91025 Évry, France

ivan.kodeli@ijs.si

 

Adjustment methods based on Bayesian inference were used to construct nuclear data covariance matrices of kinetic parameters and to adjust cross sections using the complete available information obtained from the theory and the differential and integral measurements. Performance of the methods to reduce the initial uncertainties and result in a consistent set of cross sections and covariance matrices was investigated from a rigorous mathematical point of view, taking into account the correlations among the different data and the measurements.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 612

Determination of the NPP Krško spent fuel characteristics with the Serpent and SCALE code systems

Marjan Kromar1, Bojan Kurinčič2

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

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

marjan.kromar@ijs.si

 

NPP Krško intends to transfer some of its spent fuel currently stored in the spent fuel pool to the dry storage casks. For the successful campaign a well done characterization of the spent fuel is needed. Accurate determination of the isotopic composition and specifically decay heat, activity, neutron and photon source term is namely of primary importance for a safe, secure, ecological and economical handling, transport and intermediate storage of spent nuclear fuel. In this paper a comparison of the results obtained from the stochastic neutron transport code Serpent and deterministic codes from the SCALE system is performed. Each code system has its own strengths and weaknesses. Several code options are explored to have a meaningful code comparison on one hand and to develop most suitable approach leading to best estimate values on the other hand. Since the codes from both systems rely on completely different methods, accomplished analysis provides an insight into the calculation uncertainty and indicates some recommended approaches in the spent fuel characterization process.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 613

Determination of neutron flux redistribution factors for typical PWR using Monte Carlo neutron transport methods

Tanja Goričanec1, Domen Kotnik2, Žiga Štancar2, Bor Kos2, Klemen Ambrožič2, Luka Snoj2, Marjan Kromar2

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

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

tanja.goricanec@ijs.si

 

To enable safe and continuous operation of nuclear power plant it is important to accurately control the reactivity. The reactivity in a typical pressurized water reactor is controlled by the boric acid dissolved in the moderator, control rods and burnable absorbers on fuel pellets. Control rod worth is a safety related physical parameter and can be determined by calculations and by measurements. It can be measured using different methods: e.g. boron dilution method or rod insertion method. The rod insertion method was developed at the Reactor Physics Department at the Jožef Stefan Institute. It relies on the analysis of the power signal recorded with the ex-core neutron detectors during continuous insertion of a control rod bank. The major advantage is its high execution speed (approximately 15 minutes per control rod bank) in contrast to the boron dilution method, which takes about 4 hours. During the insertion of control rod bank the spatial distribution of neutron population is changed. Since the detector measures local neutron flux at the detector location, this can lead to the non-linear power reading and control rod worth determination. To account for those redistributions neutron flux redistribution factors as a function of control rod bank axial position are introduced. In aim to calculate those factors detailed MCNP (Monte Carlo neutron transport code) model of a typical pressurized water reactor was made including detailed model of the reactor core, pressure vessel and surrounding structures, including explicitly modeled ex-core neutron detectors. The developed core model has explicitly modeled fuel rods, control rods, IFBA coating, spacer grids, upper and lower nozzles, baffle, core barrel, thermal shield, reactor vessel inner cladding and reactor vessel. The reactor core is divided into 10 axial layers and in each layer water temperature, water density, fuel isotopic composition and IFBA layer is adjusted according to the extensively validated and verified deterministic code CORD-2 output. A subroutine was developed to automate the generation of MCNP input for arbitrary fuel cycle and different parts of the cycle. A great effort was put into validation of the reactor core model by comparison to the CORD-2 power distribution calculations. It was concluded that the results for hot zero power configuration are acceptable and can be used for the out of the core calculations. Problem arises, because neutron detector is located outside the reactor core, far away from the neutron source. Neutron flux decreases for a few orders of magnitude before reaching the detector. Using Monte Carlo technique it is impossible to reach suitable statistics within a justified computer time, therefore the calculation was divided into two parts: core and ex-core calculation. The core model was used to generate the neutron source for the ex-core calculations. To speed up the neutron transport from the core to the ex-core neutron detectors, a hybrid code ADVANTG was used.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 614

A 3D MULTI-PHYSICS MODELLING OF THE TRIGA MARK II REACTOR

Christian Castagna, Carolina Introini, Antonio Cammi

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

christian.castagna@polimi.it

 

The TRIGA Mark II reactor of the University of Pavia is a research reactor that can be operated up to 250 kW in steady state. It reached its first criticality in 1965 and thereafter was used for many scientific and technical applications.
In the last years, different simulations tools were developed for a complete and accurate characterization of the reactor, analyzing in detail neutronics, burnup and thermal-hydraulics [1][2][3]. They were studied separately, without considering the mutual interaction of the different “physics” in a same simulation environment. For this reason, the present work develops a three-dimensional multi-physics approach to couple neutronics and thermal-hydraulics, in order to solve the neutron transport problem and concurrently describes the heat transfer between the fuel and the coolant at full power condition. The modelling is developed with the historical data of the first criticality configuration of the reactor, implementing the three-dimensional geometry of the active region. The effects from the coupling are important both for local (temperature and density distributions, neutron flux shape) and global parameters (multiplication factor). The results are compared with the available experimental data of the reactor criticality configurations at fresh fuel. In future works, the multi-physics approach can be extended and validated for burnup analysis.


[1] C. Castagna, D. Chiesa, A. Cammi, S. Boarin, E. Previtali, M. Sisti, M. Nastasi, A. Salvini, G. Magrotti, M. Prata, A new model with Serpent for the first criticality benchmarks of the TRIGA Mark II reactor, Annals of Nuclear Energy 113 (2018) 171 – 176.

[2] A. Cammi, M. Zanetti, D. Chiesa, M. Clemenza, S. Pozzi, E. Previtali, M. Sisti, G. Magrotti, M. Prata, A. Salvini, Characterization of the TRIGA Mark II reactor full-power steady state, Nuclear Engineering and Design 300 (2016) 308 – 321.

[3] D. Chiesa, M. Clemenza, S. Pozzi, E. Previtali, M. Sisti, D. Alloni, G. Magrotti, S. Manera, M. Prata, A. Salvini, A. Cammi, M. Zanetti, A. Sartori, Fuel burnup analysis of the TRIGA Mark II reactor at the university of pavia, Annals of Nuclear Energy 96 (2016) 270 – 276.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 615

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

Vid Merljak1, Marjan Kromar1, Bojan Kurinčič2

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

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

vid.merljak@ijs.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. Fortunately, 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.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 616

Software development for visualization of Monte Carlo results based on the MCNP program

Mario Matijević, Domagoj Markota, Davor Grgić

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

davor.grgic@fer.hr

 

MCNP is a well known and widely used computer code for general purpose Monte Carlo transport simulation of neutrons, photons and electrons through arbitrary three-dimensional configurations. Important feature of the code is graphical display of the model using X-window server, which is useful for geometry checking during input preparation. Another important graphical feature is ability to display Monte Carlo results and associated uncertainties from mesh-based tally file (i.e. meshtal file) over structured xyz mesh. Such results are quite important for the user, since they give an overall insight of Monte Carlo convergence process in phase space and effectiveness of the selected variance reduction parameters. Even though built-in routines for mesh tally plotting are useful in their basic form, there is still space left for improvement of such mesh-based visualization process. This paper presents such an attempt with development of MTV3D (Mesh Tally Visualization in 3D) program with graphical user interface for visualization of MCNP meshtal file. Basic features and functionality of the MTV3D program are presented for selected real-life shielding problems, such as PCA benchmark and PWR cask dose rates.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 618

Parametric Analysis of MCNP Multigroup Cross-Section Processing for VVER-440

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

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

branislav.vrban@stuba.sk

 

The main advantage of Monte Carlo codes lies in their ability to model complex and detail geometries without the need to accept simplifications. To achieve the best real world approximations, continuous cross section libraries (CE) are often used. These CE libraries take into account the rapid changes of XS in the resonance energy range; however, computing-intensive simulations must be performed to utilise this feature. Currently, one of the most accurate and developed stochastic MC code for particle transport simulation is MCNP, which is widely used at Institute of Nuclear and Physical Engineering FEI STU in Bratislava. To broaden our computation abilities and to allow the comparison with deterministic codes, the CE cross section library of the MCNP code is replaced by the multigroup cross section data (MG), generated by the modified versions of TRANSX and CRSRD codes. To demonstrate the new cross-section processing scheme for thermal reactor systems, the VVER-440 benchmark devoted to fuel assembly wise power distribution is selected. The parametric analysis of input options used in the cross section processing is performed to examine their impact to the investigated parameters. The obtained results are compared with continues energy MCNP and multigroup KENO-VI calculations.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Reactor Physics – 619

Measurement of the Neutron Emission Rate by the Manganese Sulphate Bath Technique

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

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

jakub.luley@stuba.sk

 

The manganese sulphate bath technique is a principal method for determination of the neutron source emission rate. This paper describes the fundamentals of this technique and an experimental instrumentation, which has been constructed at the Institute of Nuclear and Physical Engineering. The measurement system consists of the spherical shaped plexiglas bath vessel with a dry channel, where the neutron source is placed during measurement, and forced circuit with a special Marinelli beaker equipped by the 76×76 mm cylindrical NaI(Tl) gamma detector. Special attention is given to determination of the correction factors. Due to the finite dimensions of the spherical bath vessel and aqueous solution, only a small part of neutrons is absorbed by the 55Mn nuclei. The portion of neutrons absorbed on the nucleus of 55Mn is assessed by the absorption correction factor, which is calculated by Monte Carlo simulation. The circulation time of the manganese solution is also not negligible during continual measurement of the activity, therefore the sample specific activity measurement is performed from the front and the back side of the Marinelli beaker by high purity germanium detector (HPGe) and total efficiency of the NaI(Tl) detector is determined. Experimental measurements are repeated for several manganese sulphate concentrations. Obtained results are discussed and evaluated with an aim to minimize the total uncertainty of neutron source emission rate.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 704

GROWTH RATE OF RAYLEIGH-TAYLOR INSTABILITY APPLYING NANO-STRUCTURED POROUS LINING IN INERTIAL CONFINEMENT FUSION TARGET SHELLS

Arash Malekpour, Abbas Ghasemizad

Physics Department, Faculty of Science, University of Guilan, P.O.Box 41335-1914, Rasht, Iran

arash.malekpour@gmail.com

 

In this paper, the role of the nano-structured porous lining to control and to decrease of the growth rate of the Rayleigh-Taylor instability (RTI) at the ablative surface of a target shell in inertial confinement fusion has been investigated. So, the dispersion relation for the growth rate of the RTI is elicited. However the RTI happens in two stages (acceleration and deceleration phases), in this investigation this instability is assumed only in the acceleration phase applying the linear stability analysis. Learning that the porous layer absorbs the energy of the fluid and moistens the system, the growth rate of RTI for various porous linings is prospected and the porous parameter for these nano-structured porous linings are computed. It has shown that decreasing in the amount of the Bond number, causes a decrease in the growth rate, which is a function of wave number, and an increase in the amount of the permeability which increases the growth rate of RTI. Finally, it is indicated that an increase in the value of the porous parameter affords the reduction of the growth rate of RTI.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 705

Simulation of Natural Convection of Helium in DEMO Cryostat Using OpenFOAM

Rok Krpan, Matej Tekavčič, Ivo Kljenak, Boštjan Končar

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

rok.krpan@ijs.si

 

One of the hypothetical accident scenarios in DEMO fusion reactor considers ingress of helium into the cryostat due to the break of the cryogenic cooling line, which is used to cool the superconducting magnets. The accident scenario used in this study assumes that the cryostat thermal shield (CTS) remains actively cooled (e.g. approximately at a constant temperature) and the cryostat is on the outside surrounded by air at room temperature. For such scenario the majority of the heat transfer occurs in the interspace between the CTS and the cryostat inner wall. The temperature difference between the CTS and the cryostat generates natural convection of helium in the interspace, which causes a significant local cool-down of the cryostat walls. The cryostat structures in DEMO are directly connected to the bio-shield and the cryostat contraction is limited by its supports. The heat transfer and the temperature distributions on the cryostat walls is thus needed to design the supports, which must be able to withstand high temperature differences and forces caused by the cryostat contractions.
The steady-state simulations of the natural convection will be performed using the open source computational fluid dynamics code OpenFOAM. A numerical model including a 20° sector of the isolated interspace between the CTS and the cryostat inner wall will be developed. The aim of this study will be to determine the heat transfer and temperature distribution on the cryostat walls. The results will then be compared against the results obtained with the ANSYS CFX code.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 706

Comparison of MCNP and Serpent for Fusion Transport Simulations

Andrej Žohar, Žiga Štancar, Igor Lengar, Luka Snoj

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

andrej.zohar@ijs.si

 

Nuclear analysis supporting experiments, design and licencing of fusion reactors is traditionally performed using Monte Carlo particle transport code MCNP (Monte Carlo N-Particle). However, Monte Carlo code Serpent had received several enhancements that allows for its use in fusion neutronics analysis. Serpent was originally developed to be a simplified neutron transport code for reactor physics applications. Its main focus was on group constant generation with two-dimensional lattice calculations. Later the code was modified to perform Monte Carlo calculations in fission reactors and burnup calculation. With the newer enhancements of coupled neutron-photon transport for nuclear heating calculations, variance reduction methods in the form of weight windows, it also has direct equivalents to most of the surface types contained in MCNP, supports universes, cell and mesh tallies, ENDF reaction rate tally multipliers and custom response functions (e.g. flux to dose rate factors) the Serpent code has become viable for fusion neutronics analysis.
However, there are differences in implementation of particle transport between MCNP and Serpent. Unlike MCNP which uses only surface-to-surface ray-tracing method for particle transport Serpent uses both surface-to-surface ray-tracing method and the less common Woodcock delta-tracking method for particle transport which allows faster transport of particle in complex geometries by homogenizing the material total cross sections in such way that the sampled path lengths are valid over the entire geometry. This allows the random walk of particles to continue across material boundaries thus omitting the computational expensive calculation of the distance to the closest boundary surface and speeding up the particle random walk in complex geometries. However, the method also has disadvantages for fusion applications. Due to the characteristics of the described method the track length estimator cannot be used for calculation of fluxes and a less efficient collision flux estimator is used to obtain results. In vacuum regions (common in fusion models) the collision flux estimator always yields a zero result. To overcome this disadvantages of Woodcock delta-tracking method Serpent uses combination of both methods for particle transport in fusion applications.
To validate Serpent for fusion neutronics analysis a simple tokamak model will be constructed in MCNP and in Serpent with the same geometry, materials, particle source and detector positions and types. The Monte Carlo codes will be compared by comparison of calculated neutron fluxes and the efficiency of the computations with the use of parameter FOM (figure of merit) and computational times. The contribution of port openings in the tokamak structure to the neutron flux in detectors will be analysed using flagging in both Monte Carlo codes.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 707

The initial step towards JOREK integration in IMAS

Dejan Penko1, Leon Kos1, Guido Huijsmans2, Simon D. Pinches2

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

2ITER Organization, Cadarache Centre, Building 519, 13108 St. Paul lez Durance, France

dejan.penko@lecad.fs.uni-lj.si

 

JOREK [1,2] is a non-linear magnetohydrodynamic (MHD) code which was developed with the intent of producing simulations of the MHD instabilities occurring in magnetically confined plasmas. Such simulations substantially contribute to the understanding of the MHD instabilities such as edge localised modes (ELMs) and are essential for the optimization of future fusion devices such as ITER. The code itself is already well established and was validated on many occasions through simulations of MHD instabilities related to present fusion devices JET, MAST, ASDEX Upgrade, and DIII-D.

JOREK is to become another actor in the Integrated Modelling Analysis Suite (IMAS) which is being actively developed and used by the ITER organization and EUROfusion community. The list of codes integrated within IMAS workflow is gradually increasing. A few of such codes are SOLPS-ITER [3] and JINTRAC [4]. The JOREK code is meant to remain independent from IMAS while the operations in correlation to IMAS would be optional and could be used when needed. The main goal of the integration of JOREK in IMAS workflow is to set the foundations for interoperability between JOREK and the standardized IMAS databases named Interface Data Structures (IDSs). IDSs provide a uniform way of data archival and retrieval within the IMAS framework and allow to transfer data from one code component to another within larger integrated modelling workflows. In order to integrate JOREK within IMAS, therefore, it is necessary that sufficient modules and tools are provided that allow reading and writing to the relevant IDSs for its scope, namely “mhd” IDS, including its underlying Generalized Grid Description (GGD) data intended for an explicit description of the grid geometry. For that purpose, the first writing utilities were developed that extract a part of the JOREK simulation case run results, namely grid geometry and computed physical quantities for each time slice, transforms them to appropriate format and writes them to the IDS, this way archiving the data with IMAS. In this article, we present the main aspects of initial steps towards full JOREK integration in IMAS workflow that perform these functions.

REFERENCES

[1] O. Czarny and G. Huysmans. Bezier surfaces and finite elements for MHD simulations. Journal of Computational Physics, 227(16):7423 – 7445, 2008.
[2] Official JOREK website.https://www.jorek.eu/. Accessed on: 26. 04. 2019.
[3] X. Bonnin and R. Pitts et al. Presentation of the new SOLPS-ITER code package for tokamak plasma edge modelling. Plasma and Fusion Research: Regular Articles, 11(1403102), 2016.
[4] M. Romanelli and G. Corrigan et al. JINTRAC: A System of Codes for Integrated Simulation ofTokamak Scenarios.Plasma and Fusion Research, 9:3403023–3403023, 2014.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 708

INVESTIGATION OF NUCLEAR FUSION RELEVANT NEUTRON ENERGY GROUPS INFLUENCE ON ACTIVATION CHARACTERISTICS

Andrius Tidikas, Gediminas Stankunas

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

andrius.tidikas@lei.lt

 

Neutron spectrum has particular importance in nuclear fusion devices as neutrons play multiple roles in the device operation. They are the main heat carriers in the system and they are instrumental in the tritium fuel production. Furthermore, neutrons are responsible for material activation and radioactive waste production in fusion devices. They can also damage insufficiently shielded sensitive components. In order to estimate the activation of material, simulation of irradiation is being performed. Such simulation relies on material composition data, irradiation scenario and neutron spectra.
In this work, energy groups of discrete neutron spectra were examined in terms of their impact on activation inventories. Vitamin J 175 and Vitamin J+ 211 energy group structures were investigated. Two neutron spectra were selected in accordance to European DEMO nuclear fusion power plant and IFMIF-DONES irradiation facility as well as two flat spectra. Neutron spectra were obtained with MCNP neutron transport calculation code. Activation characteristics were obtained with activation inventory calculation code FISPACT. Spearman’s rank correlation coefficients for neutron spectra groups were calculated with R statistical computing software. Coefficients describe the relationship between neutron yield variation in specific energy group and activation calculation results. Three materials relevant to water cooled lithium lead (WCLL) blanket module were selected: EUROFER 97-3 reduced activation steel, tungsten alloy and PbLi. Obtained correlation coefficients correspond to aggregate neutron cross-section peaks of irradiated materials and neutron yields in energy groups. Activation characteristics include activity and decay heat values after the end of irradiation.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 709

Time Dependence Boundary Conditions During Type I ELM in ITER Scrape-Off-Layer

Ivona Vasileska

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

ivona.vasileska@lecad.fs.uni-lj.si

 

Most of the plasma edge studies are focused on the problem of the transition between a hot plasma and a material surface in tokamaks. The impurities from divertor surfaces, migrate towards the bulk plasma and can causes deviation of the parallel transport from the classical one during time. However, the transient heat loads such as ELMs (Edge-Localized modes) occur in tokamak edge during H-mode confinement lead to a significant loss of stored plasma energy. Once the ELM-driven plasma pulse has crossed the magnetic separatrix, it travels mainly parallel to the magnetic field lines and ends up hitting the divertor plate.

To keep the limits of the erosion effects, caused by the high-energy neutrals and charged particles, it is important to formulate the boundary conditions (BCs). The BCs and limiting expressions for parallel heat flux and viscosity, and their time dependencies are important tasks for plasma edge tokamaks studies.

Based on the previous article in NENE 2018, where were obtained BCs for pre-ELM during time now the aim of this work is to derive time-dependent BCs of Type I ELM state in ITER tokamak. The burning conditions of the generated plasma, that correspond to ITER are Q = 10, 15 MA baseline at q95 = 3, for which the poloidal length of the 1D SOL is ~20 m from inner to outer target. Typical upstream separatrix parameters of ne~3-5·10^(19) m^(-3), Te~100-150 eV and Ti~200-300 eV are assumed. Inclined magnetic fields at the targets of (~5o) are included, as are the particle collisions, with a total of 3.4·10^(5) poloidal grid cells giving shortening factors of 20. Secondary electron emission at the tungsten targets is neglected.

The kinetic simulations are done under under high performance conditions using the 1D3V electrostatic parallel Particle-in-Cell (PIC) code BIT1. A typical simulation requires up to 60 days running massively parallel 1152-2304 cores of the EU Marconi super-computer. The duration of the ELM pulse is taken to be between 10-20 µs. In a later stage of the work, these will be used as boundary conditions for calculations of ELM target heat loads using the SOLPS-ITER code.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 710

Shielding concept and neutronic assessment of the European DEMO Upper port

Aljaž Čufar1, Christian Vorpahl2, Christian Bachmann3, Tim Eade4, Rocco Mozzillo5

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

2Manicipal Unitary Enterprise for Waste Management “Ekores”, Setitskogo Str. 35, 220075 Minsk, Belarus

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

4Culham Centre for Fusion Energy, Abingdon, Oxon, OX14 3DB, United Kingdom

5University of Naples “Federico II”, Corso Umberto I, 40,, 80138 Napoli, Italy

aljaz.cufar@ijs.si

 

The DEMOnstration fusion power plant (DEMO) is being developed within the EUROfusion Power Plant Physics and Technology Department. Challenging aspects of this work include the integration of all required systems necessary for reactor operation, to give access during maintenance operations, while providing sufficient shielding from neutrons and gammas.
The Upper port of DEMO is challenging due to tight space constraints imposed by the toroidal field (TF) coils, a large port opening required for the tritium breeding blanket (BB) replacement sequence, and the need for various systems being integrated or passing through the port. In the present case, which is considered the most challenging, the port size and configuration are dominated by the space requirements to allow for the extraction of the BB segments and to integrate the BB coolant pipes, which are particularly large in case of helium as a coolant. In addition a plasma limiter plug is integrated inside the port. At the same time the TF coils on the exterior of the port need to be effectively shielded from neutrons to prevent loss of superconductivity due to nuclear heating and the shutdown dose rate in the cryostat needs to be within the limits. The evolution of the design and its effect on the neutronic performance are presented and the plans for future work discussed.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 711

Study of heavy ammonia production and hydrogen isotope exchange in ammonia on surfaces exposed to nitrogen/deuterium plasma

Sabina Markelj1, Matic Pečovnik1, Iztok Čadež2

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

2Retired from JSI, Jamova 39, 1000 Ljubljana, Slovenia

sabina.markelj@ijs.si

 

Nitrogen and nitrogen/hydrogen plasma have been studied and modelled for decades because of technological needs as well as for understanding of specific processes in planetary atmospheres. Nitrogen seeding of divertor plasma in tokamaks is presently used to enhance radiative cooling of the edge plasma and thus to reduce the power load to the wall and especially divertor targets. This is the reason for the more recent increase of specific research activities related to nitrogen/hydrogen plasma. Detailed kinetic scheme and data set is developed [1] to be incorporated into the EIRENE edge plasma modelling code. For this purpose, various individual collisional cross sections and rates were evaluated. Such data sets were previously tested in similar applications (e.g. [2], [3]). Even though many of the elementary atomic and molecular collision processes are well characterized there is still insufficient knowledge of surface processes under divertor relevant conditions. Beside this, the hydrogen isotope effects appear to be of main fusion plasma importance due to tritium handling of the exhaust gas from tokamak.
Here we present results of an experiment where surfaces are exposed to the deuterium and deuterium/nitrogen plasma produced in an Electron cyclotron resonance (ECR) plasma gun mounted in an UHV vacuum chamber. The ECR plasma gun is mounted in front of a polycrystalline tungsten sample and neutral atmosphere composition is monitored by mass spectrometer in real time. Experimental set-up and procedures are similar to those described in our previous publication [4] where a neutral hydrogen atom beam source was used instead of the present plasma gun (see also [5] for processes with neutral atoms/molecules). The sample is mounted on a computer controlled heater allowing temperature ramping up to 900oC in order to record the desorbing species after a specific exposure to plasma. Alternatively, we can vary sample temperature stepwise during the exposure itself and observe variation of gas composition. These are the two ways how one can differentiate processes on a relatively small sample from the processes taking place on vacuum chamber wall. Absolute production or desorption rates of individual mass species are determined by calibration of the relating particle flow to the ion current of the ion species in the mass spectrum.
Our preliminary results indicate that production of heavy ammonia (ND3) under steady exposure of tungsten to deuterium/nitrogen plasma depends on sample temperature. In our case the production of ND3 increases from 60oC to 200oC and decreases after that until our highest temperature of about 600oC. In another experiment, a sample was subjected to a neutral NH3 flow while exposed to pure deuterium plasma. In this case we observe steady decrease of production of deuterated ammonia NH2D with sample temperature (from 60oC to 350oC). Noticeable influence on mass spectra of the hot filament in ion gauge head was observed.

[1] S. Toucharda et al., Nucl. Mat. and Energy 18 (2019) 12–17.
[2] E. Carrasco et al., Phys. Chem. Chem. Phys., 2011, 13, 19561–19572.
[3] M. Sode et al, J. Appl. Phys., 117, 083303 (2015).
[4] I. Čadež, S. Markelj and M. Pečovnik, 27th Int. Conf. Nucl. Energy for New Europe, September 10-13, 2018, Portorož, Slovenia. Contr. Papers 615.1-615.8.
[5] S. Markelj, A. Založnik and I. Čadež, J. Vac. Sci. Technol. A 35 (2017) 061602-1-10.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 712

A model for stabilization of defects due to D presence during damage creation by W ion irradiation in tungsten

Matic Pečovnik1, Sabina Markelj1, Thomas Schwarz-Selinger2, Etienne A. Hodille3

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

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

3University of Helsinki, Yliopistonkatu 4, 00100 Helsinki, Finland

matic.pecovnik@ijs.si

 

The effect of D presence on the amount of displacement damage created in W during self-damaging is investigated. We have employed a macroscopic rate equation (MRE) model to analyse the results obtained in experiments where W was sequentially or simultaneously irradiated by high-energy W ions and exposed to low-energy D ions. The model includes fill-level-dependent D atom trapping in different defects and a novel damage creation, annihilation and stabilization model based on spontaneous recombination of Frenkel pairs and on stabilization of traps that are occupied by D atoms.
The MRE model was first applied on the sequential experiment where the samples were first irradiated by high energy W ions at elevated temperatures to create displacement damage and afterwards loaded with deuterium ions. Such an experiment serves as a comparison with other experiments that simulate damage evolution at high temperatures by employing room temperature damaging and subsequent damage annealing [1, 2]. This distinction in damage creation is important to consider, as displacement damage in future tokamaks will be created at elevated temperatures and not at room temperature. By simulating the experimental results, we deduced that three different defect types with several fill levels retain the majority of D in the sample. The defect densities show a linear decrease with rising temperature in the entire 300-800 K damaging temperature range with slightly different slopes for defects one and two. Defect three is independent of temperature.
The new stabilization model was developed to understand the simultaneous experiment, where damaging was done at the same temperatures as in the sequential experiment but with the addition of a simultaneous D ion exposure. To consider the effect of the W ion irradiation, kinetic de-trapping was also included as it was shown [3] that it could play an important role in the dynamics of D diffusion and trapping. A clear effect of the D presence was observed as the defect densities were higher than the values obtained in the sequential case in the region where D was present during the damaging. This effect was reproduced in our simulations by the addition of a lowered defect annihilation probability due to the presence of trapped D in the annihilating defects. The effect of D presence on additional damage created is very large when damaging at 450 K, as a two-fold increase in created defect fraction is observed, while at 800 K almost no increase is observed. This is expected as less and less defects are occupied by D with rising temperatures and therefore stabilization by trapped D becomes less and less efficient.

[1] A. Založnik, S.Markelj, T. Schwarz-Selinger, et al., Phys. Scr. T167, 014031 (2016)
[2] E. Markina, M. Mayer, A. Manhard, et al., J. Nucl. Mat. 463, 329 (2015)
[3] T. Schwarz-Selinger, J. Bauer, S. Elgeti, et al., Nucl. Mater. Energy. 17, 228 (2018)






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 713

Transport calculations for characterization of the neutron field in fusion applications

Igor Lengar, Andrej Žohar, Aljaž Čufar, Luka Snoj

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

igor.lengar@ijs.si

 

A large number of experiments involving measurements of neutron flux and spectra are performed at large tokamaks. For their optimal exploitation, i.e. to determine fusion neutron yield for example, they are supported by radiation transport calculations. The large majority of transport calculations are performed with the Monte Carlo code MCNP, which is the reference code for analyses of several of the larger tokamaks including the Joint European Torus (JET) and planned tokamaks ITER and DEMO.

In case of the larger tokamaks a vital problem of neutron flux calculations outside the vacuum vessel, where also the majority of neutron detectors is located, is the large attenuation of the flux from the plasma source to detector positions, associated with larger uncertainty in the neutron flux. This is reflected by the fact that calculation results frequently do not agree with measurements. One of the reasons for this discrepancies is the necessary simplification of calculation models since the complexity of the real device is impossible to implement into models. These include geometry simplification and material homogenisation.
The correctness of calculations depends to a large extent on the correct amount of materials, present in the models. Moreover the most important are the materials, with a high probability to influence the neutron transport, usually materials with large cross sections (scattering, absorption, … )

In the present work neutron transport with the MCNP code is studied in tokamak geometry with the emphasis to find the most important isotopes, which influence the results of neutron transport up to the exterior of the vacuum vessel and the detector locations. The information is important for optimisation of the MCNP models – it can be used in order to take special care of the important materials and their geometrical distribution while homogenising the less important materials or isotopes. This information is apriori not available from MCNP and a procedure has been developed in order to extract it from the raw MCNP output. In addition such analyses gives good insight into the characteristics of the neutron transport inside tokamaks.

The calculations are studied on a tokamak model with the correct material mix as found in the first wall, vessel and biological shield of large tokamaks. MCNP calculations are performed and the most important isotopes influencing the neutron fluxes at positions of ex-vessel detectors, identified. Neutron pathways are studied and the achievable accuracy of results estimated.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 714

The effect of mesh refinement on the calculation of cryostat thermal loads in DEMO during the incident helium ingress

Martin Draksler, Boštjan Končar, Matej Tekavčič

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

martin.draksler@ijs.si

 

During helium or air ingress accident into the DEMO cryostat, the natural convection is established inside the cryostat, which causes cooling or heating of the cryostat inner wall. For DEMO the cryostat walls are envisaged as a thin-walled structure, where contraction is constrained by its supports to the bioshield. These supports can be designed by knowing the local distribution of the temperature and heat transfer coefficient (HTC) on the cryostat walls. The natural convection is stronger in event of He ingress, thus resulting in higher HTC values than in the case of air ingress.

The present study considers the He ingress event during the leakage of cryogenic lines while the cooling of the magnet system remains active to represent the most severe scenario. A steady-state CFX simulation was performed using the simplified DEMO tokamak model to assess the thermal loads on passive thermal shields and cryostat walls. The predicted temperature distribution in cryostat walls as well as the estimated HTC depend strongly on the density of numerical mesh for fluid domain, especially its near wall resolution at fluid/solid interfaces. The analysis with six different numerical meshes has been conducted. Only a sufficiently refined mesh in the fluid near-wall region leads to the mesh-independent prediction of temperatures in the cryostat walls. This however requires a huge computational effort.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Fusion – 715

Kinetic properties of SOL with intermittent filamentary transport

Jernej Kovačič1, Stefan Costea2, Tomaz Gyergyek1, Inaki Gomez4, Tsviatko Popov5, Roman Schrittwieser2, Codrina Ioniţă2

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

2University of Innsbruck Institut of Ion Physics Plasma Department, Techniker str. 25, A-6020 Innsbruck, Austria

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

4Faculty of Physics, St. Kliment Ohridski University of Sofia, 5 James Boulcher Blvd., 1164 Sofia, Bulgaria

jernej.kovacic@ijs.si

 

Tokamaks seem to be currently the most viable design for the construction of a fusion power plant. They combine relatively simple design with good performance. There are however still open issues that affect its performance, with the exhaust of the core plasma being one of the main. The tokamak exhaust, the scrape-off-layer (SOL), has open field lines, which drive the undesired products, i.e. helium ash and residual heat, into the divertor region. Due to the nature of transport across the magnetic field lines and the complexity of the transport parallel to the magnetic field, the SOL is very difficult to model and subsequently predict its behaviour.
Here we are presenting a new approach to the modelling of the SOL, by introducing an intermittent filamentary transport into domain and a fully kinetic development of the SOL in parallel direction. We do so using a fully-kinetic particle-in-cell simulation code BIT1. The code is able to simulate the full length of a magnetic flux tube in the SOL with suitable boundary conditions. We have upgraded the code, so that particle and energy source is capable of mimicking the intermittent filamentary transport. We have also improved the radial diffusion module, so that it now not only capable of locally and globally ambipolar diffusion, but of locally non-ambipolar, globally ambipolar diffusion.
With the newly upgraded code we have simulated a train of blobs arriving into an empty simulation domain and building a SOL from scratch, preserving its kinetic properties through development of the particle energy distribution functions in time. From the simulation results we have been able to observe the development of the SOL kinetic properties and construct boundary conditions for the fluid approach from the first principles of the kinetic simulations.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radiation and Environmental Protection – 804

MACCS2 statistical analysis to evaluate the off-sites consequences of a severe accident at the Krško NPP for Emergency Preparedness and Response purposes

Antonio Guglielmelli, Antonio Cervone, Federico Rocchi

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

antonio.guglielmelli@enea.it

 

ENEA, with the aim to increase its competencies in the field of Emergency Preparedness and Response (EP&R), has recently enlarged its set of simulation tools with the Melcor Accident Consequence Code System version 2 (MACCS2) within the US-NRC CSARP framework. MAACS2 has been developed by the Sandia National Laboratories to perform risk assessment studies of potential off-site consequences on the public and environment of a radioactive material atmospheric release due to severe reactor accidents. The need to enhance the ENEA simulation capabilities is due to the fact that Italy is surrounded, at less than 200 km from the borders, by several foreign NPPs for which it is necessary to improve the level of preparedness to have a preliminary idea of the degree of radiological impact over the Italian regions for a severe accident. WinMAACS3.11.2, a user-friendly Windows Interface version of the MACCS2 code, has been used to provide, by means of a simplified selection of input parameters, the results of the post-processing of several MACCS2 calculations. MACCS2, which was designed primarily as a probabilistic risk assessment tool (PRA – level 3), can sample hourly annual weather data and generate statistics that describe the effects of weather conditions at the time of a release. MACCS2 includes the entire dose-relevant pathways and evaluates land contamination areas, doses to individuals and populations, health effects and risks, and economic losses resulting from an accident. MACCS2 code has been therefore adopted to perform a preliminary statistical radiological impact study of a severe accident at the Krško NPP, which is one of the closest NPPs to the Italian borders. In order to achieve a site-related calculation, some specific data, as Source Term (ST) and meteorological dataset, were defined. The ST values chosen for the simulation were 1.0E+16 Bq for Cs-137 and 1.0E+17 Bq for I-131 with a dynamics of “PUFF” type (i.e., 1 hour of emission). The hourly meteorological data, which include wind direction, velocity, stability class, and rain intensity, were collected from the History+ service of Meteoblue weather data. The simulation domain has been modeled with a polar-grid coordinate system subdivided in 32 compass directions up to a radial distance of 140 km from the emission source with the aim to evaluate the radiological impact on a polar grid element that accurately identifies the Italian territory. The characteristics of land that surrounds Krško NPP were also taken into account defining the land fraction for each grid element of the MACCS2 spatial domain. The probability of exceeding a series of values of total ground deposition and time-integrated air concentration over the simulation domain was evaluated. The consequences to the population in terms of Total Effective Dose Equivalent (TEDE) and of thyroid dose were also investigated.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radiation and Environmental Protection – 805

Radioactivity in the Environment: SNSA’s New Web Portal

Michel Cindro, Samo Tomažič

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

samo.tomazic@gov.si

 

The Slovenian Nuclear Safety Administration (SNSA) has been collecting on-line data from gamma dose rate probes since the first one was installed shortly after the Chernobyl accident. What started as a single Geiger-Muller counter connected to the computer has been growing and today the SNSA manages more than 60 gamma dose rate probes, specialized aerosol monitors that can detect alpha, beta and gamma contamination in the air as well as stationary spectrometers that can automatically measure and identify radioactive elements in wet and dry deposition.
More than one decade ago, SNSA developed its very first comprehensive system for data collection, aggregation and presentation for expert and public use. The primary function of the Early Warning System (EWS) was alarming of the SNSA staff in case of elevated levels of radiation on the territory of Slovenia. Additional features of EWS were informing the public of current state of radioactivity levels in the country and educating interested parties on radiation protection and radiation monitoring through supplementary materials made freely available. During the long-time usage of EWS many new lessons were learned, therefore, we decided to totally rebuild the web portal. The new and modern EWS, now called Radioactivity in the Environment (RVO), is much more intuitive, more reliable, easier to administer, and has many new features. To name just a few that stand out the most: better statistical representation of data, quick review of current radiation monitoring situation, coordination of mobile units during emergencies (including Android and iOS mobile app), simulation of emergencies and merging it with prediction software in order to take full advantage of the system for exercise and education purposes as well as to enhance decision making in a nuclear or radiological accident. The SNSA has in the past developed a database of off-line laboratory measurements of radioactivity in environmental samples dating from 1960s, called Database of the Radioactivity in the Environment (ROKO), which has now been incorporated into the RVO solution. In case of an accident, this integration will provide easily accessible reference values for the comparison of historical data with values obtained from mobile units.
Building such a software takes a lot of knowledge, time and financial resources. In this paper, we will focus on all aspects of developing and deploying such a system, and the challenges we had while trying to upload it to the governmental cloud services and set up a publicly accessible portal. In addition to that, we will briefly discuss plans for the future and the need to keep up with the developments in the field of radiation measurements and advances in data transfer technologies. Once in its final production state, RVO will help us to better manage everyday situations as well as emergencies, and on the other side, it will help the public to keep track of the current radiation monitoring situation in the country in real-time.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radiation and Environmental Protection – 807

Feedbacks from Radiation Protection Courses in Nuclear Training Centre

Tomaž Skobe, Matjaž Koželj

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

tomaz.skobe@ijs.si

 

The paper will present experiences, good practices and feedbacks from radiation protection courses, organized at Nuclear Training Centre Ljubljana. Nuclear Training Centre has been also certified according to ISO 9001 quality standard since December 2006 and last year a new standard ISO 9001:2015 has been adopted. Nuclear Training Centre is a part of Jožef Stefan Institute which is authorised institution in the field of radiation protection and radiation protection training in Slovenia. A wide spectrum of courses for different users is regularly organised. The content of courses is different for various type of exposed workers and the courses are divided to three major fields: radiation protection for industrial and other practices, radiation protection for medicine and veterinary medicine and radiation protection for nuclear facilities. Participants are requested to answer the evaluation questionnaire at the end of all courses. All comments are then distributed to lecturers and the plan of necessary improvements is made. In the paper feedbacks from courses performed in last 10 years are collected and analysed.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radiation and Environmental Protection – 808

Transposition of BSS Directive in practice: Slovenian results for stakeholder engagement in medicine, emergency management and indoor radon

Nadja Železnik

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

nadja.zeleznik@eimv.si

 

EURATOM 2013 Basic Safety Standard (BSS) Directive was transposed into the EU national legal frameworks due to February 2018. The BSS directive establishes uniform basic safety standards for the protection of the health of individuals subject to occupational, medical and public exposures against the dangers arising from ionising radiation. It applies to any planned, existing and emergency situation which involves a risk from exposure to ionising radiation. One important aspect which is investigated in Horizon 2020 ENGAGE (ENhancinG stAkeholder participation in the GovernancE of radiological risks for improved radiation protection and informed decision-making) project is to identify and address how the BSS requirements for stakeholder engagement are arranged in three fields of exposure to ionising radiation: medical use of ionising radiation, emergency planning and response (EP&R) and indoor radon.
In ENGAGE participating countries the stakeholder engagement was investigated from several points of view: first, the frameworks and rationales for stakeholder engagement in radiation protection and the related legal or contextual drivers was investigated, and secondly, the research on how stakeholder engagement in radiation protection issues is enacted in practice and what its impact is on radiation protection decision-making was done. The investigations followed the adopted methodology and included the analysis of what justify and/or prescribe stakeholder participation related to radiation protection and how these prescriptions and expectations are enacted in practice. For Slovenia, the analysis included beside the legal framework also three case studies: i.) EP&R information and communication in practice at different levels (national and local), ii.) Justification, optimisation, education and training at Institute of Oncology Ljubljana, and iii.) Stakeholder engagement in case of national radon action program. The paper will present the findings of the research.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radiation and Environmental Protection – 809

Design of a protective container for fuel-containing materials

Serhii Kupriianchuk1, Marjan Kromar2

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

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

s.kupriianchuk@ispnpp.kiev.ua

 

The main idea of the research is to develop the design of a container for nuclear waste resulting from the accident at the 4th Unit of the Chornobyl nuclear power plant in 1986. The container shall provide safety functions during transportation and storage of fuel-containing material resulting from the accident at the Chernobyl Nuclear Power Plant.
Since structure, geometry, physical, chemical and radiation characteristics are not so widely explored, as for example spent nuclear fuel, implementation of safety criteria for the handling and storage of containers dedicated to fuel-containing material is not a trivial task.
The design of the container was based on the basic safety criteria. These assumptions were used to develop input data for a 2D model of a typical RBMK-1000 reactor fuel cell. The main input data of the model are the geometric dimensions of the elements, the concentration of isotopes 235U, 236U, 234U and 238U, the specific density and temperature of the fuel cell components, irradiation time and cooling time. The model was used to deplete fuel and to determine the isotopic compositions with the TRITON-NEWT module from the SCALE package.
For the final cooling phase ORIGEN module was used to determine the isotopic composition of spent nuclear fuel, the change in the activity of elements over time, the intensity and energy distribution of the photon and neutron source.
The obtained isotope composition data and its characteristics were implemented to create a simple model of protective container suitable for the MCNP code version 5. This model represents a source (corium) placed inside a container made of stainless steel. As a result of calculations, the dose rate on the container surface is determined, taking into account appropriate photon and neutron fluxes. In addition, a sensitivity study is performed, where the dose rates are calculated based on the thickness of the container and the distance from the container surface.
This simulation was conducted to obtain preliminary conservative results that would allow to improve the container design and would help to develop a more sophisticated and effective model of protective container.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radiation and Environmental Protection – 810

Nuclear emergency preparedness – Short overview of EU approaches

Miodrag Stručić1, Luca Ammirabile2, Juan Carlos De La Rosa Blul3, Patricia Pla Freixa4, Montserrat Marin Ferrer2, Miguel Angel Hernandez Ceballos5

1Joint Research Centre of the European Commission, Westerduinweg 3, 1755 ZG Petten, Netherlands

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

3Joint Research Centre of EC, Institute for Energy, Westerduinweg 3, 1755 ZG Petten, Netherlands

4Joint Research Centre of the EC, Petten, Westerduinweg 3, 1755 LE Petten, Netherlands

5EC, Directorate-General Joint Research Centre Institute for Energy Safety of Present Nuclear Reactors Unit (SPNR) Plant Operation Safety, PO Box 2, 1755 ZG Petten, Netherlands

miodrag.strucic@ec.europa.eu

 

Shock caused by Fukushima Daiichi accident is still echoing in Nuclear Society. Lack of adequate estimation of tsunami dreadful effects and other demonstrated phenomena remind us that, beside scientific and professional expertise in domain of nuclear accidents, we have to consider many other aspects to be better prepared.
International Atomic Energy Agency (IAEA) with main objective to support member countries in nuclear and radiological safety, based on the Convention on Early Notification of a Nuclear Accident and the Convention on Assistance in Case of a Nuclear Accident or Radiological Emergency, provides different means for support in emergency situation. Nevertheless, during emergency, European Commission (EC) have important role in providing support to its member states, but also closely cooperate with IAEA through established networks.
There are different level’s efforts to share operating experience from Nuclear Power Plants (NPP) events, but experience from disasters is far bigger in other industries and could be examined and adopted to nuclear. Though, handling nuclear specific issues in NPP accidents are only a small part of a full scale approach to disaster. One of the latest efforts to extract different industrial sector’s valuable lessons and practice from Emergency Preparedness and Response for use in nuclear sector resulted in new Nuclear Energy Agency and the Organisation for Economic Co-operation and Development (NEA-OECD) report, contributed with results of some EC Joint Research Centre (JRC) in-depth accident analyses.
Although it seems that recent activities in EU are holistically aimed to “all-hazard” approach to emergency preparedness and response, this paper gives short overview of different industrial sectors approaches to disasters with their possible adaptation in nuclear emergency preparedness strategy.

Keywords: Emergency Preparedness, Joint Research Centre, disaster, cross-border risk, all-hazard approach






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radioactive Waste Management – 904

The TN MW®, a flexible solution for Waste Management

Assia Talbi1, Renaud Leblevennec2, Catherine Grandhomme1

1ORANO, 1 Place Jean Millier, 92400 Courbevoie, France

2ORANO TN, 1 rue des Hérons, 78180 Montigny le Bretonneux, France

3Unknown Organization, Unknown, Unknown, Slovenia

assia.talbi@orano.group

 

Nuclear operations induce a large variety of waste to be managed by operators. These waste significantly vary in terms of type, volume and activity, therefore multiplying packing, storage and transportation solutions to be implemented.
To address this issue, Orano developed a new cask design named the TN MW® to provide operators with an all-in-one solution, dedicated to waste conditioning, transportation, storage and disposal.
This highly flexible cask aims to simplify legacy, operational and D&D waste management while optimizing associated costs. Thanks to its reduced dimensions and weight, the TN MW® is particularly adapted to the management of small quantities of various waste and allows easy handling operations.
Orano’s 50 years’ experience as a casks provider allow operators to have access to a flexible and cost effective waste management solution.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radioactive Waste Management – 906

Impact of fuel type and discharge burnup on source term

Pauli Juutilainen, Silja Häkkinen

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

pauli.juutilainen@vtt.fi

 

Knowledge of spent nuclear fuel (SNF) source term (decay heat, reactivity, nuclide inventory, other relevant properties of SNF) is essential in the safe handling and final disposal of SNF. For example, decay heat power is one of the key limiting factors to determine how densely the fuel canisters can be packed in the final disposal tunnels. The fuel type and burnup affect the nuclide inventory of the SNF and therefore have an essential impact on the source term. Fuel discharge burnup has increased over time and new types of fuel typically enable higher burnups than before.

In Finland, two different reactor types have been operated (VVER-440 and BWR) and two others (EPR and VVER-1200) are planned to be operated in the near future. The fuel assembly design used in these types of reactors varies in many parameters such as e.g. geometric shape, axial enrichment zoning and nuclide content (VVER-1200). Also, several different fuel assembly types can be used in one type of a reactor. The possible differences include e.g. average enrichment, enrichment zoning, number of burnable absorber rods and geometric parameters, such as fuel pellet dimensions and cladding thickness. The operating parameters such as power and boron concentration also depend on the reactor and fuel assembly types and have an effect on the SNF source term.

In this work, the source term of different VVER-440 and VVER-1200 fuel assemblies are calculated as a function of fuel burnup using the Monte Carlo particle transport code Serpent 2. The effect of burnup and fuel type to different components of the source term such as decay heat, activity and the nuclide inventory will be examined.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radioactive Waste Management – 907

Liquid Waste Treatment of a Hydrazine Based Reductive Metal Ion Decontamination (HyBRID) Process

Hui-Jun Won1, Hee-Chul Eun1, Jeikwon Moon2, Seon-Byoung Kim2, Bum Kyoung Seo1, Sang Yoon (Andrew) Park3

1Multiple organizations possible, Unknown, Unknown, Slovenia

2Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, 305-353 Daejeon, South Korea

3KAERI (Korea Atomic Energy Research Institute) Decommissioning Technology Research Division, 989-111, Daedeok-daero, Yuseong-gu, 305-353 Daejeon, Republic of Korea 34057, South Korea

nsypark@kaeri.re.kr

 

Organic acids such as oxalic acid, citric acid, nitrilotriacetic acid (NTA), and ethylene-diaminetetraacetic acid (EDTA), which are used for chemical decontamination processes of a nuclear power plant (NPP), generate chelating ligands that are often problematic for the safety of radioactive wastes from the decontamination processes. In particular, EDTA, which is used in the commercial decontamination process, forms a complex with radionuclides during the decontamination process, and this complex is not easily decomposed and increases the mobility of the radionuclides. To resolve this problem, the Korea Atomic Energy Research Institute (KAERI) developed a new decontamination technique consisting of a hydrazine-based reductive metal ion decontamination (HyBRID) process and a related pre-oxidative process. The chemical agents used in the HyBRID process are N2H4, H2SO4, and Cu2+ ion. Sulfuric acid (H2SO4) and potassium permanganate (KMnO4) are used in the pre-oxidative step. The combination of the processes, which is called the SP-HyBRID process, is a new decontamination technique for the decommissioning of NPPs. This SP-HyBRID process has many advantages. First, the stability of waste disposal is remarkably improved because there is no organic or complexing agent in the decontamination solution. Second, a precipitation process is used for the decontamination waste treatment instead of an ion exchange process, and this contributes to a large reduction of radioactive waste generation. There are a considerable concentration of hydrazine and many kinds of metallic cations containing radioactive materials (Mn2+, Fe2+, Ni2+, Cr3+, Cu2+, Co-60, Co-58, Mn-54, and Cr-51) in the SP-HyBRID decontamination liquid waste. The hydrazine is decomposed using hydrogen peroxide (H2O2). All of the metallic cations in the liquid waste tend to co-precipitate with BaSO4 precipitates during a precipitation process using Ba(OH)2, and they are mostly removed from the decontamination solution by means of a filtration process. Therefore, the co-precipitation characteristics of the metallic ions play an important role in the purification performance of the liquid waste. Our study focused on the purification properties of the liquid waste. For this reason, the chemical conversions of various metallic cations and sulfate ions in the liquid waste were simulated with adding Ba(OH)2 during the precipitation process. In addition, co-precipitation characteristics of radioactive materials with BaSO4 precipitates were evaluated. Our results confirmed that SP-HyBRID decontamination liquid waste could be purified to a satisfactory level by the precipitation process using Ba(OH)2 without the need for an ion exchange process.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radioactive Waste Management – 908

Setting the Authorized Limits of the Discharges from the Restored Disposal Sites of Žirovski Vrh Uranium Mine (Slovenia)

Metka Kralj1, Milko Janez Križman2, Mitja Eržen1

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

2Retired, nn, nn, Hungary

metka.kralj@arao.si

 

ABSTRACT
The authorized release of radioactive gaseous and liquid discharges to the environment is part of the radioactive waste management of the nuclear or radiation facility. Exposure of the public is controlled by monitoring of emissions and of the environment. The regulatory authority sets the discharge and dose limits, which are based on the public dose limitation resulting from an optimisation process. These are usually one or two orders of magnitude lower than the general radiation exposure limit for the public.
In Slovenia, the authorized limits are imposed for the operating nuclear and radiation facilities (NPP, research reactor, some hospitals). The only exception in this respect was the uranium mine at Žirovski Vrh. A general dose limit of 1 mSv/year was used during the time of its operation while the authorized limits for disposal sites for uranium mining and milling residues were set only in the post operational period (in 1996). These limits, including the authorized dose limit of 0.3 mSv/year, were based mainly on the actual results with a reasonable expectation that the then levels would be reduced after restoration works. Monitoring in the transitional five-year period from the end of remediation works to administrative closure of the site confirmed that the performance of the remediated site was in accordance with the authorized limits.
The new legislation, based on EU Directive 2013/59/Euratom, requires that the authorized limit values (in terms of concentrations and/or annual discharges) are set at the limitations of the effective dose for the public. This was not the case for the U-mine Žirovski Vrh where the discharges are hard to be regulated in comparison with other operating facilities, equipped with the relevant technical possibilities.
The paper presents the revision of the proposed authorized limits including their derivation, which was initiated by the Slovenian Agency of Radwaste Management. These limits cover radon exhalation, liquid discharges and external radiation at the mine waste disposal site Jazbec and liquid discharges from the mine galleries. They are based primarily on the original total authorized effective dose limit for the public and partial doses arising from the existing exposure pathways, taking into account also the results of the post-closure monitoring.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Radioactive Waste Management – 909

Determination of Disposal Density of VVER-1200 Spent Fuel Loaded Canisters in Horizontal Geological Disposal

Gürel Özeşme1, Banu Bulut Acar2

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

2Hacettepe University, Beytepe Çankaya/Ankara, 6400 Ankara, Turkey

gurelozesme@itu.edu.tr

 

Geological disposal is the most accepted method for permanent disposal of spent nuclear fuel and high-level waste. There are various geological disposal concepts under development in many countries and these concepts have differences mainly in the geometry and material of disposal canisters, geological formations of host rock and orientation (vertical and horizontal) of disposal canisters. The aim of this study is to determine a relationship between geological disposal density (the amount of radioactive waste that can be safely emplaced per unit area of the geological repository) and spent nuclear fuel storage period before geological disposal, a number of canisters emplaced into a canister and thermal load of spent nuclear fuel assemblies. These canisters are simulated as to be loaded with spent nuclear fuel assemblies discharged from VVER-1200 reactors and disposed horizontally in the granitic rock formation. The ANSYS Fluent 19.2 finite element code is utilized to determine geological disposal densities. In the first part of the study, mentioned reactors spent fuel characteristics (amount, isotopic composition, heat generation rate, etc.) are evaluated for normal operation expected burnup values by using the MONTEBURNS 2.0 code. Then, 3-dimensional ANSYS model of geological repository consisting horizontally disposed canisters, buffer material and host rock regions is developed and transient thermal analysis is performed. As a result of the study, geological disposal densities of canisters loaded with spent fuel assemblies are determined by taking into account the thermal constraints (90oC at canister surface). Thermal analysis are repeated for disposal canisters with various number of spent fuel assembly sets (per canister) for evaluating the effect to waste disposal densities.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1004

Safety issues and the role of corporate level in nuclear power industry

Silviu Gabriel Hada, Dan Serbanescu, Adrian Jelev

National Company “NUCLEAR ELECTRICA” s.a., 33 Blvd Gh. Magheru, P.O.Box 22-102, 70164 BUCHAREST 1, Romania

dserbanescu@nuclearelectrica.ro

 

One of the outcomes of the post Fukushima review on safety was to enhance and improve the participation of the nuclear power companies in the process of safety review for their plants. For this type of activity, the terminology of safety oversight at corporate level is used and the activity is conceived as an addition and a supervision of the existing safety reviews already in place.
In compliance with this trend, the regulatory environment was reviewed at national level in Romania, in accordance with the best practices defined by the international organizations, by including the specific requirement that the nuclear power companies have to perform safety oversight review at the corporate level.
The paper presents the main elements of the implementation of the national requirements at the corporate level in the National Nuclear Power Company – Nuclearelectrica S.A.
The main elements follow the best practices, which are independent on the reactor type, as defined in internationally recognized recommendations, as for instance the WANO or INPO documents.
On the other side, there are other specific features, as for instance the ones related to the plant type (in our case a CANDU type), involving a close correlation with the best practice in the plant type groups and company specific features and procedural systems.
The paper presents the existing framework of the safety oversight at corporate level in our company, significant aspects of the current results on the use of the new mechanism and future trends and steps for its consolidation






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1005

Extension of TRANSURANUS fuel behavior modelling with uncertainty propagation and sensitivity analysis

Zsolt Soti

JRC-Directorate for Nuclear Safety and Security , Hermann-von-Helmholtz-Platz 1, P.O. Box 2340,, 76125 Karlsruhe, Germany

zsolt.soti@ec.europa.eu

 

The imperfect knowledge of input variables and model parameters causes uncertainties in deterministic fuel rod simulations. Those uncertainties influence the safety and design features of the nuclear fuel rods. The TRANSURANUS fuel performance code, using the built-in Monte Carlo method, can simulate random uncertainties for up to 70 different input variables, including also time-dependent input quantities. For each random set of input variables, necessary for one run of the code, 45 different time-dependent outputs are saved. In this way the built-in Monte Carlo module shortens the time necessary for calculations and makes it possible to run TRUNSURANUS several thousand times for one probabilistic input setup. The number of simulations and the input parameter distributions are defined in a single input file.
We have extended the TRANSURANUS postprocessor package with Python modules that are able to perform simple statistical analyses of the output and input variable distributions. The TUPython package has a graphical user interface and allows quantifying the time-dependent input and output uncertainties and plotting the probability distributions of the variables. The time-dependent Pearson’s and Spearman’s correlation coefficients can be calculated and visualized for input-output (I-O) combinations chosen by the user. With that the sensitivity of a chosen output against a given input can be quantified.
Data export in csv file format is realized in TUPython. For each case, the files with row statistical I-O data of all related Monte-Carlo runs can be saved and uploaded for the further mathematical analysis in different statistical software packages like DAKOTA, Excel, etc…
This work shows the features of the TUPython packages, the graphical user interface and presents a case study for uncertainty and sensitivity analysis of a nuclear fuel rod behavior simulation.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1006

Uncertainties Assessment and Sensitivity Analysis of the CIP0-1 RIA Test by means of TRANSURANUS

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

 

Increasing the discharge burn-up of used fuel pursues the objective of improving the economic performance of nuclear energy. As a consequence, safety criteria should be carefully reconsidered to take into account the more demanding operating conditions occurring at high burn-up. For these reasons, the analysis of accident conditions (RIA, LOCA) at high burn-up is challenging and gives rise to the need for a re-assessment and an accurate validation of fuel performance codes.

The Working Group on Fuel Safety (WGFS) of the OECD/NEA has promoted since 2011 a benchmark of fuel codes focused on RIA calculations. A new phase of the benchmark started in March 2018 and its final report is underway. The main objective of this initiative is to perform a statistical analysis of codes’ results. In particular, the impact of the initial state and key models (including burn-up effects) on transient results has been investigated by means of uncertainty assessment and sensitivity analysis. The case study selected by the participants is the CIP0-1 experiment which was performed within the framework of the CABRI International Programme (CIP). The peak burn-up of this test is about 75 MWd/kgHM. Experimental measurements during transient and PIE analyses give the opportunity to discuss codes’ predictions in detail.

This paper presents an overview of the TRANSURANUS calculations performed in the frame the OECD/NEA RIA benchmark Phase III. The assessment of uncertainties and the analysis of sensitivity (Spearman Rank Correlation Coefficients) were carried out by means of the TRANSURANUS modelling resources and its extended statistic version based on the Monte Carlo method (also presented at this conference). Results are used as a valuable source of information to identify, in agreement with the outcomes of other international benchmarks (e.g., FUMAC), priorities for its future development. In addition, it is expected that these results could be of help in the definition of a best practise for the use of TRANSURANUS in this kind of statistical studies.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1007

COMPUTER SIMULATION OF NUCLEAR REACTOR ON SPHERICAL STANDING BURNING WAVE

Yurii Leleko1, Volodymyr Gann2, Ganna Gann2

1National 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

2National Science Center ”Kharkov Institute of Physics and Technology”, 1, Academicheskaya str., Kharkov 61108, Ukraine

makswell.com@gmail.com

 

Concept of the traveling wave nuclear reactor (TWR) is one of the brilliant ideas of 20-th century. It suggests using depleted uranium (or thorium) as fuel and promises to supply inexhaustible source of energy worldwide. This idea was proposed by S.M. Feinberg, realized theoretically by L.P. Feoktistov and developed in many publications, in which several ways of its practical implementation were suggested. One of the most promising designs (TWR) is a fast reactor with negative reactivity feedback, which is able to work in manoeuvrable mode. In the most of published works, various cases of linear traveling burning waves in stationary neutron-multiplying media were considered. Another idea assumed to use standing burning waves in the reactor of the cylindrical symmetry with periodic shuffling fuel elements This requires the development of a theory of a nuclear burning wave in a moving medium. In this article, the theory of nuclear reactor on spherical standing burning wave is developed. The neutron kinetics of a nuclear burning wave in a moving neutron-multiplying medium in the presence of nuclear reactions was developed.
Computer simulation of moving and standing spherical burning wave in a nuclear reactor was performed. The fuel moved with acceleration towards the origin with the velocity V (r) = VR (R / r)2 and spherical burning wave travelled away from the origin. Burn-up material was removed from the centre of reactor, and depleted uranium continuously enters into the peripheral area. The reactor’s control elements provided its criticality and negative reactivity feedback. A standing nuclear burning wave can exist in such a system. Non-fissile isotope 238U was converted to fissionable 239Pu in standing burning wave.
Mathematical simulation of such reactor was carried out using MCNPX code. This model is a ball with a radius of 1.2 m, filled with fuel based on uranium dioxide. In the traveling spherical wave mode, nuclear burning begins in the central region of the core containing enriched uranium. When the concentration of Pu-239 in U-238 becomes quite high due to 239Pu production according to the scheme U-238 + n = U-239 › Np-239 › Pu-239, then a spherical burning wave appears, it breaks away from the ignition region and continues the radial movement to the edges of the core during 60 years. In our model calculations, the burning wave velocity was ~1 cm/year at a power of 250 MW.
A computer simulation of the spherical standing nuclear burning wave was carried out. The burning wave was at rest relative to the reactor vessel when velocity of the medium amounted ~1 cm/year. Computer simulation results were in agreement with theoretical dependencies.
Possibility of using depleted uranium as a nuclear fuel in reactors on spherical burning wave is confirmed.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1008

Analysis of IFA-507 by means of the TRANSURANUS Code

Rolando Calabrese

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

rolando.calabrese@enea.it

 

TRANSURANUS is a widespread and well-known fuel performance code. Its capabilities are continuously refined and improved by the team of developers (JRC/ITU, Karlsruhe). In parallel, the set of experiments employed for validation purposes is enlarged with additional experimental data. The IFA-507 test was designed and conducted to study the performance of thermocouple-bearing rods under a rapid (<5 s) load follow/power shock. The IFA-507 experiment was performed in the OECD Halden Reactor. This experimental test was included in the set of experiments proposed within the FUMEX-II Research Project coordinated by the IAEA. The experimental data is available in the International Fuel Performance Experiments Database (IFPE).
The six rod cluster in IFA-507 was surrounded by a moveable silver shield encased in stainless steel which is capable of reducing the power to about 50% of the full power rating. A rapid withdrawal of the shield induces fast power transients in the fuel pins. The fuel temperature of two rods have been monitored during the transient (TF3, TF5). The burn-up of these rods was about 18 MWd/kgUO2. A preliminary analysis has been carried out aiming at testing the performance of a mechanistic fission gas release model that has been recently implemented in the code (version v1m1j18). Preliminary results are discussed in comparison with the experimental data and predictions of the conventional fission gas release model. In addition, results are reviewed in the light of most recent developments of the physics-based fission gas release model.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1009

INFLUENCE OF LOAD FOLLOWING ADJUSTMENT IN THE ELECTRIC POWER SYSTEM ON THE OPERATING AND ECONOMY OF NEW NPP KRŠKO 2

Tomaž Ploj, Samo Fürst

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

tomaz.ploj@gen-energija.si

 

The electric power system is a complex infrastructure of national importance that requires the maintenance of existing power facilities and the development and investment in new power facilities. In this context, it is extremely important that due to the growing integration of distributed renewable energy sources in the system, which have influence on bigger units that have to adjust their power in the electric power system. There is still prevailing misleading opinion also among energy professionals that nuclear power plants, unlike other technologies, are rigid and inflexible and can not be used for load following like thermal or gas power plants.

The paper presents the results of an external study developed for the investor of the proposed new Nuclear Power Plant (NPP) Krško 2. The aim was to verify the effects of the power adjustment in electric power system on the operation, as well as on the economy of new NPP Krško 2 based on the known electricity market models and prices. In addition, the requirements for new builds and the practice in other countries, compared with other technologies, will be presented.

Results shows that the required operational characteristics of the nuclear power plant are at the level of the most modern conventional power plants, also the generation 2., and are capable of providing the most complex system adjustments required by network. That is why the new NPP Krško 2 will be maximal flexible, but still economically optimal in base load production.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1010

Method for Coolant Velocity Estimation in the Reactor Core Using In-Core Rh SPND detectors

Sándor Kiss, Sándor Lipcsei

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

lipcsei.sandor@energia.mta.hu

 

Inhomogeneities of the coolant passing through the reactor core induce small fluctuations in the neutron flux in the steady state of the reactor, which causes small transients in the in-core neutron detector signals with a time delay proportional to the distance between the detectors. A challenge in the determination of the time delay comes from the global component of the neutron flux fluctuations virtually obscuring the component induced locally by the perturbations passing by the detectors. Hence a correlation method is used to determine the time delay and the coolant velocity from it.
This paper shows the components of the neutron noise signals and their separation from each other through a model of perturbations passing through the reactor core. The coolant velocity estimation based on this model is demonstrated using real measurements of SPND signals from a VVER-440 reactor.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1012

Why a Typical Event in a Nuclear Power Plant Can Lead to a Major Nuclear Accident – Probability, Causes and Consequences

Venceslav Gospodinov Kostadinov

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

kostadinov.venceslav@gov.si

 

The article describes the original views and qualitative and quantitative estimates of the likelihood that a typical nuclear event in a nuclear power plant can turn into a major and more serious nuclear accident.

The possible causes that the initial event in the nuclear power plant goes into greater nuclear accident are described and any possible consequences are predicted.

A new quantitative mathematical model for probability and risk estimates for nuclear power plants is given and described.

Some sample examples of calculations are presented with a new model for probability and risk calculations. A sample calculation is provided for the Fukushima nuclear power plant. A Japanese power station is shown and possible causes for the occurrence of such a serious accident, some consequences of this accident, and a quantitative calculation of the risk with a new model are shown. Described and quantified are the possible probable causes that such a serious nuclear accident has developed from the accident.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1013

Progression of Station Blackout Event in PWR plant

Andrija Volkanovski, Andrej Prošek

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

andrija.volkanovski@ijs.si

 

The progression of the event in the nuclear power plant and corresponding consequences depend on the number of parameters and corresponding uncertainties. The goal of this study is identify, classify and analyse the main sources of uncertainties of the parameters that affect the progression and consequences of one selected event for the nuclear power plant.
The selected event for the purposes of this study is Station Blackout (SBO) event resulting in loss of all alternate current power sources in the nuclear power plant. This event was selected as one of the most demanding events for the light water reactors.
The identification of the most important parameters was done on basis of results of the previous parametric studies and sensitivity analysis of deterministic calculations.
Identified most important parameters and corresponding uncertainties were classified in two categories: external and internal. The internal parameters were defined as those parameters that indicate the state of the primary coolant system of the nuclear power plant (and secondary system in case of pressurized water reactors). All other parameters were classified as external.
The analysis of the uncertainties of the selected parameters shows that dominant contribution to the progression of the event and final consequences, for SBO event, have operator actions (especially recovery of system safety functions). The remaining parameters have small/negligible impact on the event progression so they can be omitted in further analyses.
Based on the above analysis the SBO event progression tree is developed with main events and operator actions that are expected to be considered in further analyses, for example with the Bayesian belief network, of extended SBO.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1104

Public Opinion about Nuclear Energy – Year 2019 Poll

Radko Istenič1, Igor Jenčič2

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

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

radko.istenic@ijs.si

 

The Information Centre which is part of the Nuclear Training Centre at the Jožef Stefan Institute informs the visitors about nuclear power and nuclear technology, about radioactivity and about Krško Nuclear Power Plant.
Our main target population are the schoolchildren from the 8th and 9th grade of elementary school with their teachers (in total close to 8000 per year). The visitors can choose between live lectures on nuclear technologies (fission and fusion), a lecture about use of radiation in medicine, industry and science and a lecture on stable isotopes. For younger visitors, a lecture about energy and an energy workshop is available. The visit includes a demonstration of radioactivity and a guided tour of a permanent exhibition.
We monitor the opinion trends since 1993 by polling about 1000 youngsters every year. To obtain their opinion based on the knowledge from everyday life we conduct the poll before the youngsters listen to the lecture or visit the exhibition. We will present and comment trends over the last 26 years.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1105

Safety Consideration on Decommissioning Strategies of Nuclear Power Plants in Korea

Kyung-Woo Choi

KINS (Korea Institute of Nuclear Safety), 62 Gwahak-ro, Yuseong-gu, 34142, DAEJEON, South Korea

kwchoi@kins.re.kr

 

The legal framework of decommissioning in Korea was improved through the revision of Nuclear Safety Act and its subordinate status in 2015. The technical standards and criteria related to the inspection on the decommissioning nuclear facilities and site release condition were established subsequently. According to the revised NSA, the preliminary decommissioning plan (PDP) including decommissioning strategy should be submitted as one of conditions for approval of NPP at the construction/commissioning stage. After periodical review of PDP every ten years, the NPP operator should submit the final decommissioning plan (FDP) for decommissioning approval within 5 years after permanent shut-down of NPP. The decommissioning strategy of nuclear facilities was established at construction stage for preparation of decommissioning and reviewed for the decommissioning approval.
The first PWR-type NPP in Korea, Kori unit-1, was decided to shut-down permanently in 2017 and expect to be decommissioned. If the existing NPPs do not permitted to extend the operation, more than ten units of NPP is expected to be permanently shut-downed and decommissioned in 10 years. The Korean regulatory body have experienced for decommissioning of small-sized nuclear facilities like research reactors and nuclear fuel-cycle facility. On the other hand, we have no regulatory experience on NPP decommissioning. The decommissioning of Kori unit-1 is the first case for large commercial nuclear facility like NPP in Korea. With respect to the safety regulation on Kori NPP decommissioning, the detailed technical standards and safety review guideline for FDP have to be developed and applied for Kori NPP decommissioning. In order to prepare the detailed safety guideline, a regulatory R&D project was launched in 2016 and is currently in progress on the basis of national decommissioning roadmap.
The establishment of decommissioning strategy (DS) is one of the requirements for decommissioning approval and the regulatory body should review the justification of selected strategy. The safety consideration factors for set up the DS are as follows; availability of radioactive waste management facilities, impacts on near operating units in multi-facility site, end state after site remediation, financial assurance followed by cost estimation, availability of dismantling/decontamination technologies and human resources, and so on. The DS should suggest the decommissioning method and schedule also. The results of safety assessment and environmental impact assessment caused by decommissioning activities as well as lessons learned from domestic/international decommissioning experiences should be reflected at the selected decommissioning strategy and methods. Considering all above the safety factors related to selection of DS, the regulatory body should review the adequacy and feasibility in terms of DS justification through FDP reviewing process.
In this presentation, the status of Korean regulatory framework and Kori NPP’s decommissioning strategy as well as current operator’s decommissioning preparation activities will be introduced. In order to set up the regulatory position on justifying the selected decommissioning, the related safety consideration factors will be analyzed and discussed in safety regulation aspects either.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1106

Radiation Protection Training for Radiation Protection Culture: What Can and What Can Not Be Done

Vesna Slapar Borišek, Matjaž Koželj

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

vesna.slapar-borisek@ijs.si

 

While the term “safety culture” is well known among professionals in nuclear technology, people outside nuclear facilities are not familiar with the meaning and the implementation of the approaches based on safety culture. This applies to radiation practices in general industry, services, research, education, and medical institutions. In nuclear facilities safety culture influences all aspects of management, planning, production and maintenance with radiation protection being only one of the “targets” of safety culture that accompanies the first and initial target, namely nuclear safety. Considering the positive influence on radiation protection implementation, we can say that the safety culture has become one of the most important keystones of radiation safety in nuclear facilities.
Unfortunately, it is not true for other practices which we have mentioned before. Neither legal requirements for radiation safety, nor common occupational safety initiates and disseminates implementation of safety culture in radiation protection in the sense that we have learned from nuclear technology. Additional problem is that radiation workers and other professionals involved in radiation practice do not know exactly what safety culture is, what should be done to implement it, and what should be achieved.
To clarify and present the importance of safety culture to radiation protection professionals IRPA has prepared special document IRPA Guiding Principles for Establishing a Radiation Protection Culture, where use of the term “radiation protection culture” is recommended for radiation protection elements in safety culture. It is our aim to discuss different elements of radiation protection culture in non-nuclear organisations and to analyse which of these could be addressed during (individual) training and to present some solutions for the compensation of deficiencies which are the result of this limited approach to improvement of radiation protection culture.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1107

ESSENTIAL EDUCATIONAL ACTIVITIES FOR PUBLIC AS A PART OF IMPLEMENTATION OF THE CROATIAN STAKEHOLDER INVOLVEMENT PROGRAMME RELATED TO RADIOACTIVE WASTE MANAGEMENT

Želimir Veinović1, Ivica Prlić2

1University of Zagreb Faculty of Mining, Geology and Petroleum Engineering, Pierottijeva 6, 10000 Zagreb, Croatia

2Coronel Institute for Occupational and Environmental Health Academic Medical Center, P.O. Box 22700, NL-1100 DE AMSTERDAM, Netherlands

zelimir.veinovic@rgn.hr

 

Croatian radioactive waste management program is in its most vulnerable phase, considering urgent need for confirmation of acceptability of the only remaining potential location for the low and intermediate level radioactive waste (LILW) storage facility. After the site selection process (which is not a scope of this paper) and approval by the Government, the Čerkezovac site, at Trgovska gora, is the one remaining potential location for the LILW storage facility in Croatia.
Specific circumstances of Slovenian and Croatian co-ownership of Krško nuclear power plant (KNPP) and urgent need for the removal of excess LILW from KNPPs storage facility, due to overfilling, are rushing Croatian regulator and Programme implementor to a fast practical solution acceptable to a local population. Local community is opposing the idea of the LILW storage facility in their region, and the Croatian state border neighbouring communities of the Republic Bosnia and Herzegovina are against it as well. These counteracting activities, based on poor, sometimes not at all scientific arguments, exist mainly due to the decades long gap in distributing the core information and almost non existing education of Croatian population which includes most of involved stakeholders.
The lack of public information/education is becoming one of the main obstacles for the Croatian LILW management program. In this paper the most relevant and most urgent activities for the Croatian stakeholder involvement are being presented and discussed in order to direct it to a Programme. Research relating to radiological protection relevant for the site and its inhabitants should be conceived of as transdisciplinary and inclusive, integrating citizens, scientists and other stakeholders’ input into research (Perko T., et al 2019, J. Radiol. Prot.). It is to be performed prompt and in qualitatively acceptable form harmonized with the new and modern European contours of a Strategic Research Agenda for Social Sciences and Humanities in radiological protection.

Radioactive waste, stakeholders, Programme, Croatia, LILW, radiological protection, responsible research and innovation






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1108

EFFECTIVE NUCLEAR LITERACY

Lina Alkawass, Saleh Ismail

Atominstitut, Schüttelstr.115, A-1020 Wien, Austria

lina.alkawass@tuwien.ac.at

 

The global energy system is meeting major challenges by reducing fossil fuels-based energy production and expanding the transition towards sustainable and low carbon technologies. However, nuclear energy, which could play a key role as a baseload technology while reducing greenhouse gas emissions, is still burdened by the dilemma of radiation phobia and Chernobyl-Fukushima accidents. It is remarkable that many countries, including those which are using nuclear power, are not able until now to build a rational opinion regarding the nuclear and radiation technology through media, educational and outreach programs. It is, therefore, necessary to understand the underlying factors influencing the energy culture by applying an effective bottom-up procedure. To facilitate the longitudinal studies of the attitudes and energy knowledge, a dynamic web-based nuclear energy literacy questionnaire (NELQ) has been developed. The NELQ consists of two modules; the first is general and suitable for the public and schools’ students, while the second includes additional technical topics which can be added to measure the energy knowledge of young professionals. This paper discusses the results of testing the NELQ as well as evaluates the activities of some schoolteachers after they had attended a two-week training program at the TU-Wien.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1109

A Virtual Reality System (VRS) for RAPID

Valerio Mascolino1, Alireza Haghighat1, Nicholas Polys2

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

2Multiple organizations possible, Unknown, Unknown, Slovenia

val@vt.edu

 

This work discusses a Virtual Reality System for RAPID (VRS-RAPID) web-application that allows for fast input preparation, output visualization and interactive collaboration between users of the RAPID Code System. RAPID is capable of analyzing and monitoring nuclear systems such as spent nuclear fuel (SNF) pools and casks, and reactor cores.
RAPID is based on the Multi-stage Response-function Transport (MRT) methodology, that partitions a problem into a series of independent stages, based on the problem physics, that are then coupled via response functions of coefficients. RAPID calculates prompt and delayed fission sources (steady-state and time-dependent), eigenvalue (keff), kinetic parameters (e.g., Rossi-alpha), subcritical multiplication factor and detector responses. The response functions for different stages are pre-calculated for a wide range of system parameters (e.g., burnup and cooling time for SNF systems) and are then coupled in real-time via a linear system of equations. This allows for analyzing any configuration of the specific system in real-time (seconds) by combining and interpolating the coefficients’ database.
VRS-RAPID allows users to generate RAPID inputs online using a series of interfaces that closely represent the geometry of the systems to be analyzed, eliminating the need of compiling long textual input files. In addition, various users can collaborate on the same model at the same time. Various interfaces are present, e.g., for loading of fuel assemblies in SNF systems or for inserting control rods within TRIGA reactors, and new ones can be easily developed based on the phenomena of interest.
Once the input has been prepared, VRS-RAPID executes the RAPID Code System on its server. The VRS provides estimates for running times and live timers. Once the RAPID execution is completed, the output is shown in the VRS’s output elements. The distribution of fission neutrons is shown by means of an X3D model. The model can be rotated and moved by the users connected to the VRS at that moment. Normalization and scales of the X3D model can be changed by the user. 2D plots at different axial locations and axially integrated can be activated by the user. System eigenvalue and subcritical multiplication factor are shown in a box next to the X3D model. RAPID’s output files can be downloaded for further studies as well.
VRS-RAPID is also equipped with an inspection algorithm that allows for comparison of calculated and experimental (as input by the user) detector responses at different positions. The algorithm can identify fuel diversion and/or misplacement by comparing the detector responses using the Maximum Likelihood – Expectation Maximization (ML-EM) technique. This capability provides unique in-situ monitoring for nuclear safeguards applications and reactor operation.
A new feature in the VRS allows for it to be used in “presenter mode”, as opposed to the “collaborative mode” described so far. When this feature is active, only one user can actually build and modify the RAPID input. However outputs and results are visualized simultaneously by all the other users. This feature makes the VRS a useful tool for interactive learning in university classes and/or for professionals training. The system allows seamless change between the two modes. Alternatively, some “viewer” users can be selectively inhibited/allowed to build a model and running it by leaving the other users’ privileges unaltered. These features allow for a great flexibility for academic classes, professional training, and presentations.
VRS-RAPID can be accessed via any web-browser by authorized users. This eliminates the need of having the RAPID Code System on one’s computer, and removes any need for maintenance of the code system. Every user action is recorded on the server in a log for quality assurance and control purposes.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1110

Training of Radiation Protection Officers in Slovenia: What are We Missing?

Matjaž Koželj, Vesna Slapar Borišek

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

matjaz.kozelj@ijs.si

 

The current system of radiation protection in Slovenia was designed with the introduction of Ionising Radiation Protection and Nuclear Safety Act in the year 2002, which was approved during the harmonisation of Slovenian legislation with the EU regulations. Secondary legislation which was approved in the following two years also provided particularities of radiation protection implementation and training of exposed workers in nuclear facilities, medical and other organisations. While the organisation and training of exposed workers in nuclear facilities have not been modified considerably from the previous period, the organisation of radiation protection in other organisations has been redefined and elaborated in details. One of the novelties related to the organisation of radiation protection was the introduction of persons responsible for radiation protection and authorised radiation protection experts.
Persons responsible for radiation protection are de facto radiation protection officers. Staff members of radiation protection units in nuclear facilities also belong to this category, but these units were established at the time of Krško NPP commissioning and most of these officers were trained extensively either in Vinča Institute in former Yugoslavia, or in Oak Ridge, USA, and their training (in the past) has been accepted without special additional conditions. The reason was probably the transfer of general radiation protection arrangements to the Krško NPP together with the technology and legal requirements for nuclear safety from the US legislation.
It was not the case with the persons responsible for radiation protection. Their duties were initially related mostly to licensing or registration of practices, and later to implementation of radiation protection measures. They should take care of all necessary arrangements that radiation protection measures are implemented, but their capacity to be practically involved in radiation protection tasks or supervise them was limited. The main reason was that their training was limited, and being the person responsible for radiation protection is usually their supplemental duty on the job. New Slovenian legislation based on updated EU BSS from the year 2013 imposes even more technical and practical duties to persons responsible for radiation protection, but instead of revision and expansion of training, it was cancelled and we have finished with radiation protection officers who are, in many cases, less proficient in radiation protection than regular radiation workers.
In our contribution, we will discuss the deficiencies of the current approach to the training of persons responsible for radiation protection in Slovenia and the need to reintroduce and expand the training of these persons. We will also compare our approach with the practices in some other countries.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1111

Representative social survey on public energy preferences, renewable, fossil and nuclear alternatives

Börcsök Endre1, Zoltán Ferencz2, Veronika Groma2, Szabina Török1

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

2Institute of Nuclear Research of HAS, P.O. Box 51, H-4001 Debrecen, Hungary

 

torok.szabina@energia.mta.hu

 

In the past energy policy decision-making was handled as a single criterion decision-making approach either concentrating on economic, supply security or environmental issues etc. Such approach is not suitable for present policy maker because it cannot handle the complexity of currant social, technological, environmental, and economic factors and there is no chance to consider trade-offs of several conflicting aspects. Multicriteria methods (MCDA) provide a flexible tool that is able to handle and bring together a broad range of principles in different ways and thus offer valuable assistance to the decision maker.
Two different approaches ranking future energy technologies were compared. Participants of the study were first asked to provide their subjective weights of criteria. To weigh criteria for expert can be straightforward but it does not represent the public opinion. Previous studies showed that survey of lay man needs a multistep approach since the criteria are not familiar to them and represent very different phenomena depending on education and experience with energy use. A countrywide questionnaire survey was conducted and answered by 1000 people. The sample was representative for age groups, gender, educational level and types of settlements. Indicators and alternatives were selected based on the focus group interviews results and a complex questionnaire was assembled for 32 indicators and 8 energy alternatives. Questionnaire data were evaluated by analytical network process (ANP) that is used in complex decision processes by e.g. ecomomists.
Based on aggregated results environmental criteria and renewable energy as alternative are highly preferred, however by splitting the survey answers according to educational level and types of settlements will provide opportunity to monitor the energy preferences of different social groups.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1112

Regulatory Aspects Regarding Topical Peer Review on Ageing Management Program

Jure Škodlar

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

jure.skodlar@gov.si

 

Slovenia participates in the Topical Peer Review (TPR) on aging management under the Euratom Nuclear Safety Directive. The Slovenian Nuclear Safety Administration (SNSA) has been involved first in all steps of the preparation of the WENRA Technical Specifications, which was prepared by all participating countries.
The SNSA than prepared Slovenian report in cooperation with the Krško NPP, focusing on parts related to regulatory oversight and assessment. The SNSA has than continued with activities under the scope of TPR with a peer review of all participating countries reports and review meeting in Luxembourg. The final stage is preparation of country-specific action plans. In 2019 the Action Plan will be submitted to the ENSREG and after that, the status of the implementation of the actions will have to be reported each two years.
In our case based on the WENRA Technical Specifications, the Krško NPP ageing management program was taken into account for the TPR. The Krško NPP intends to extend its operation beyond its original design life to 60 years and established aging management program (AMP) was one of the prerequisites for prolongation. The SNSA after approving their AMP, follows the all AMP processes in the Krško NPP by performing thematic inspections, oversight on the plant outages, through periodic safety reviews, modifications, review and assessment, operating experience feedback follow up, review and assessment of AMP documentation and other activities.
The paper focus on TPR and challenges, which were shown during TPR process related to the Krško NPP ageing management of equipment important to safety, such as electrical cables, concealed piping, reactor pressure vessel and concrete containment structures from the regulatory point of view. The AMP requirements and implementation for these area will be presented. The SNSA activities regarding AMP in the Krško NPP will be described as well.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Regulatory Issues, Sustainability and Education – 1113

A Role of Radiation Protection Officers and Radiation Protection Experts

Helena Janžekovič

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

helena.janzekovic@gov.si

 

The main component assuring radiation safety is knowledge. The Council Directive 2013/59/Euratom from 2013 emphasises a role of knowledge management. Looking to lifetime phases of facilities and activities with radiation sources, e.g. NPPs, knowledge management related to nuclear and radiation safety issues is needed e.g. in project organizations, environmental expert companies, designers, building companies, suppliers, operators, radwaste management agencies, as well as rescue teams and regulatory authorities. Organisations mentioned should manage appropriate level of knowledge among their staff. This might be a challenging task in particular in countries with fast ageing population of workforce in nuclear industry, e.g. in EU as noted by ENEN.
As radiation safety is not included in typical education scheme, additional trainings and education as appropriate are needed. Such education and training are specific for a specific task or job and are well established for workers within such facilities. However, knowledge of workers is not enough to assure safe operation of facilities using radiation sources. To cope with a large scope of knowledge which might be required two concepts were developed as given in Council Directive 2013/59/Euratom and IAEA GSR Part 3, i.e. concepts of:
• radiation protection officer (RPO)
• radiation protection expert (RPE).
Both concepts were introduced to emphasise the difference in knowledge between workers in facilities and those with specific role there requiring not only broader but also deeper understanding of radiation protection.
The concepts have been developed for decades. However, it seems that a role of RPOs is well established. But there is a lot of models how to introduce a role of QEs in regulatory framework. Already in 1998 EC published Communication 98/C 133/03 with basic syllabus of training for QEs in radiation protection. Today their certification, authorization or recognition is not harmonised, e.g. only few years ago IRPA published the Guidance on Certification of a Radiation Protection Expert. Within the EU specific projects have been conducted in order to achieve mutual recognition of QEs. EUTERP has been established.
The article analyses a role of QEs in countries where they are already strongly involved in regulatory processes, such as a role of QEs in Slovenia, as well as their role where they are introduced for the first time. In some countries, e.g. USA, a concept of QE is not present. In particular, difference between a role of QEs and RPOs in industry is analysed. The article focuses on a role of QEs regarding the requirements given in IAEA GSR Part 4 (Rev 1) and guides given in IAEA GSG-13 (2018).
Moreover, the article gives a practical example of a role of qualified experts involved in assuring safety in industrial radiography as one practices with relatively challenging safety records, i.e. reports on accidents and incidents in this practice are regularly reported by the IAEA. The article aims to help the countries with developing regulatory framework for radiations safety to analyse implementation of a concept of QEs and to help upgrading efficiency and effectiveness of such framework in countries where a role of QEs has been already introduced.






10.09.2019 17:30 Regulatory Issues, Sustainability and Education

Regulatory Issues, Sustainability and Education – 1114

Roadmap to a new research reactor in Slovenia

Jan Malec1, Anže Pungerčič1, Bor Kos1, Klemen Ambrožič1, Andrej Žohar1, Vladimir Radulović1, Anže Jazbec2, Sebastjan Rupnik2, Vid Merljak1, Aljaž Čufar1, Žiga Štancar1, Luka Snoj1

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

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

jan.malec@ijs.si

 

As of 2019, four research reactors are planned to be built in the European Union and Switzerland according to the IAEA database of research reactors and out of those four only one is already under construction. None of the reactors currently planned or in construction is flexible enough to be adapted to a wide array of different experiments or suitable for education activities. On the other hand, 35 research reactors are currently being decommissioned in Europe. Since the nuclear power production is not expected to be phased out in the near future, European institutes, companies, and universities are already running short on facilities dedicated to research and education needed to support the research and development in the field of nuclear technologies.
The Jožef Stefan Institute (JSI) in Slovenia has been operating a TRIGA Mark II research reactor [1] since 1966. The reactor has been used for training and education, development of detectors, production of medical isotopes and research. In addition, the activities connected to the research reactor have supported the Krško NPP which resulted in significant economic savings and reduced dependence on third parties. In 2019, the reactor is fully utilized and involved in multiple international projects. After more than 50 years of successful operation, we see a need for a new research reactor, which will allow Slovenia to advance nuclear research, keep the nuclear expertise and maintain the safety culture in the next 50 years. The shortage of research reactors in Europe creates a unique opportunity for Slovenia to become a European center for research activities connected to nuclear reactors.
In this paper will present an overview of current state of research reactors, reactors currently under construction and in decommissioning. We will investigate the factors that will influence the reactor design, which are: the technology used in European power reactors, the needs of large-scale nuclear research projects in Europe, which are Myrrha, Jules Horowitz Reactor and the PALLAS and the needs of other potential users. In addition, we will discuss how to design a reactor that will cover a vast arrange of currently identified uses and be flexible enough to support various physics experiments in the future. The findings of the paper will serve as a basis for a feasibility study conducted as the next step in the reactor acquisition [2].
[1] SNOJ, Luka, SMODIŠ, Borut. 45 years of TRIGA Mark II in Slovenia. V: JENČIČ, Igor (ur.). Proceedings, 20th International Conference Nuclear Energy for New Europe 2011, September 12-15, 2011, Bovec, Slovenia
[2] Feasibility Study Preparation for New Research Reactor Programmes, IAEA Nuclear Energy Series No. NG-T-3.18, Vienna, 2018






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Materials, Integrity and Plant Life Management – 301

Status of materials for nuclear technologies in Ukraine

Victor Voyevodin

National Science Center ”Kharkov Institute of Physics and Technology”, 1, Academicheskaya str., Kharkov 61108, Ukraine

voyev@kipt.kharkov.ua

 

Now nuclear energy meets the best the principals of sustainable development, one of the main demands of which is the presence of sufficient fuel-energetic resources on their stable consumption into the long-term prospective.
Nuclear Power Plants now are the main source of electrical and thermal energy and guarantees the energetic independence of Ukraine. In some months of 2017 – 2018 years about 60% of electric energy was produced by 15 working NPP.
Unfortunately in spite of considerable efforts of scientists in many countries of the world which develop the nuclear power the economically necessary levels of operation on working reactors are not attainable.
It is caused by insufficient radiation resistance of main structural materials of existing nuclear plants, namely, of stainless steels of different classes and of zirconium alloys
At this time the behavior of materials for nuclear energetic is the critical point which mainly limits the safety and commercial advantage of nuclear power plants
The role of structural materials of the core and reactor vessels consists in the providing of minimal consequences in an emergency, and, in other words, in solving of main problems of the reactor safety and economy, also including questions of waste management .
The presented paper describes the modern problems of radiation behaviour of structural materials in conditions of reactor operation and also possible in reactors of future generations
Now thermal-neutron reactors, pressurized water reactors or boiling reactors are the base of the world nuclear power (WWER-440, WWER-1000, PWR, BWR).The main components of thermal reactors that are subjected to the intensive radiation exposure are the pressure vessels, fuel elements claddings, pressure vessels internals, which can lead to degradation of original physical- mechanical properties and dimensional changes.
Almost in all countries of the world new NPP units are oriented towards the reaching of service life not less than 60 years with possibility to life extension by 20-30%, that is, to 80-90 years. The service life of reactor vessel mainly determines the service life of NPP units. Understanding of the mechanisms of radiation embrittlement of vessels will allow forecast this radiation phenomenon on their further operation.
The accident on Japan NPP “Fukushima” demands the increase of the safety and economy of zirconium alloys by formation of prescribed structural states and surface modification. The use of protective coatings is prospective and economically advantageous solution because it is possible to use existing zirconium claddings and technologies of their production.
The actual way of investigation is analysis and generalization of mechanisms of influence of different operational factors on swelling and microstructure of austenitic steels, prediction of serviceability of elements of pressure vessel internals of reactor units at high damaging doses.
Realization of presumptuous programs for development of reactors of IV generation and fusion reactor needs the solution of the problem of development of principally new radiation-resistant structural materials which will operate in the core of fast reactors (E>0,1 MeV) up to damaging doses 200 displacements per atom and temperatures of irradiation 370-710 C.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Materials, Integrity and Plant Life Management – 309

High Fidelity Computation Models to Calculate the Effective Material Properties of Porous Cells

Nima Fathi

University of New Mexico, Department of Mechanical Engineering MSC01 1150 1 University of New Mexico, Albuquerque, NM, USA-New Mexico

nfathi@unm.edu

 

Our efforts are to develop and demonstrate high fidelity models to calculate effective material properties for porous materials. The initial area of concentration is on nuclear fuels. Most nuclear fuels do not achieve theoretical density when manufactured. Therefore, properties of interest such as heat capacity and thermal conductivity will depend on the porous structure of the fuel. This research concentrates on stand-alone models of nuclear fuels until its utility is demonstrated. Then the appropriate models will be incorporated into the MOOSE code system. However, with the advent of 3D printing and other modern manufacturing techniques, porosity in manufactured materials is likely to be an issue in many other areas besides nuclear fuels. The potential for future programmatic funding will probably expand beyond the development of nuclear fuels. Here the recent computational results and verification and validation assessment on the effective thermal conductivity calculation will be presented.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Research Reactors – 508

Computational-Experimental Investigation of Fission Thermal Probe

Nima Fathi

University of New Mexico, Department of Mechanical Engineering MSC01 1150 1 University of New Mexico, Albuquerque, NM, USA-New Mexico

nfathi@unm.edu

 

In Transient Reactor Test (TREAT) facility the magnitude of nuclear heating is one of the most crucial attributes in test design, analysis, and data interpretation. Some steady state tests are able to measure heating directly via enthalpy rise of coolant in thermally-isolated loops, but an enormous number of important tests must rely upon nuclear modeling to correlate core operating parameters to specimen nuclear heating. Uncertainties innate to these models and their inputs can prevent experimenters from achieving desired conditions and hamper advanced models from describing the phenomena to the level of confidence needed. The situation can be even more difficult for non-steady-state tests (i.e. transient tests) where nuclear heating is intentionally varied over short time scales. This research develops a novel nuclear heating sensor technology which directly measures the parameters of interest using spatially-resolved real-time thermometry of fissionable instrument materials, demonstrate the instruments’ use in TREAT, and perform data comparisons to nuclear models to facilitate an advanced understanding of transient power coupling. Here we present our recent results of the designed probe and its testing outcome.






10.09.2019 15:40 “Poster Session I” and 11.09.2019 10:10 “Poster Session II”

Nuclear Power Plant Operation and New Reactor Technologies – 1002

FEASIBILITY OF USING ERBIUM AS BURNABLE POISON IN VVER-1000 FUEL ASSEMBLY

Jiří Závorka1, Martin Lovecky2, Radek Skoda1

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

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

jiri.zavorka@centrum.cz

 

This study shows the feasibility of using Erbium as a burnable absorber (BA) in VVER-1000 reactor. Nowadays, Gd2O3 is commonly used BA in the form of additive compounds in the nuclear fuel for light water reactors. This analysis compares standard gadolinium absorber with a variant of absorber made of erbium in several forms. It is focused on the reactivity decrease and relative power of fresh fuel assemblies. The main aim of this study is to investigate neutronics properties of Erbium as the burnable neutron absorber and the possibility of the use in VVER-1000. Base on the Serpent calculations, a new type of fuel with Erbium absorbers are designed.






11.09.2019 15:40 Poster Session I

Radiation and Environmental Protection – 801

Continuous Maintenance of Early Warning System

Miha Blažič

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

miha.blazic@gov.si

 

Continuous Maintenance of Early Warning System

Miha Blažič, Dušan Peteh, Samo Tomažič
Slovenian Nuclear Safety Administration, Litostrojska cesta 54, 1000 Ljubljana, Slovenia
miha.blazic@gov.si

The Monitoring section, at the SNSA, continuously monitors external radioactivity levels in the environment over the territory of Republic of Slovenia through measurement devices and communication lines. Probes are located across whole country and around Krško NPP. There are 72 probes, out of those, 16 are located near Krško NPP and it is called Early Warning System or simply EWS.

Probes also measure potential radioactivity released into the environment by the former uranium mine at Žirovski Vrh, the TRIGA Research Reactor and the Central Storage near Ljubljana for Radioactive Waste. The goal of radioactivity monitoring in the environment is to timely detect any elevated levels of gamma dose radiation in the environment and the general radioactive contamination.

The success of accurate data on radioactivity and working probes depends solely on regular maintenance. At SNSA there are two principles that are implemented for maintenance of the EWS. First is fast response, and second is professionalism and awareness of the purpose of fully functioning of the EWS. On every three years probes are sent for calibration. Currently EWS uses almost two decades old measurement technology. Maintenance, of either probes or measurement device, is carried out by SNSA technical trained personnel. During the maintenance, the field technical personnel is always in contact with personnel at SNSA to monitor data on EWS software.

The paper will discuss and provide an outline of importance of maintenance of the EWS and how it works on field. It will also outline the new planned update that will replace existing probes with modern ones that are better and technologically more advanced. The main characteristic of new EWS will be a two-way communication which will enable remote configuration, meaning, for example, that transfer of data, in case of radiological or nuclear disaster, will be shortened, instead of 30 minutes to 5 minutes.






10.09.2019 14:40 Thermal-hydraulics

Thermal-hydraulics – 201

Constraining input uncertainty sources of PSA by sensitivity analysis using FFTBM-SM

Andrej Prošek, Andrija Volkanovski

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

andrej.prosek@ijs.si

 

There are multiple sources of uncertainties in the probabilistic safety assessment (PSA) of event (e.g. station blackout). The sources of the uncertainty in the PSA are therefore minimized before their integration within PSA. The purpose of this study is to propose and demonstrate the fast Fourier transform based method by signal mirroring (FFTBM-SM) for judging the impact of given parameter uncertainty on selected plant parameter. The identification of the most important input parameters gives opportunity to constrain the uncertainties before PSA analysis is performed.
The plant parameters and corresponding uncertainties are classified into two main categories, external and internal parameters. The external parameters are all those parameters characterizing/affecting the progression of the event that are not related to the status of the primary and secondary system of the PWR. The internal parameters include all parameters that characterise the state of the primary and secondary system.
The proposed method is applied on the station blackout (SBO) event of the two-loop pressurized water reactor (PWR) with utilization of the RELAP5/MOD3.3 Patch 04 computer code for the simulation of the SBO scenarios. The impact of SBO to emergency diesel generator operation time and time of cooling restoration through secondary side, both classified as external uncertain parameter is analysed. Different types of primary system coolant loss scenarios (existence of normal system leakage, seal and letdown loss) and primary system depressurization strategy, identified as the main internal input uncertain parameters, are evaluated and obtained results are presented.
Obtained results demonstrate that larger source of uncertainties are external (i.e. operator actions) than internal parameters. The operator actions provide strategies to restore the electric power and with that the function of core cooling and by this preventing the core damage. Such strategies largely influence the accident progression, much more than internal parameters uncertainties, or uncertainties in the initial and boundary conditions, and the code model uncertainties. In the context of deterministic calculations to support PSA, the uncertainty quantification of initial and boundary conditions and code model uncertainties present smaller contribution to the results on one side and require huge effort on the other side.






10.09.2019 15:00 Thermal-hydraulics

Thermal-hydraulics – 202

Analysis of statistical uncertainties in a direct numerical simulation of flow in a backward facing step

Jure Oder, Iztok Tiselj

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

jure.oder@ijs.si

 

In this paper, we present the statistical uncertainties in the results of direct numerical simulation of thermal and fluid fluctuations in a flow of liquid metal past a backward-facing step (BFS) with finite dimensions and solid walls. Since the turbulent flow has chaotic temporal oscillations, the flow is compared to experiments and other tools by using statistical quantities. However, such reductions introduce uncertainties that are reduced with the number of steps and length of simulation. We analysed the statistical convergence of our simulation in around 50 chosen points through approximately 10 million time steps or around 5000 dimensionless time units.

Simulations are performed with the NEK5000 code. The most notable feature of this code is the use of spectral elements to solve for velocity, temperature and any other passive scalar. It is an open source code developed by the Argonne National Laboratory.

Spectral element method is a hybrid method between finite element method and a collocation spectral method. The method divides the computational domain into finite elements, within which a spectral method is used to solve for variables. This method allows for the use of spectral method in irregularly shaped geometries and to perform direct numerical simulations in such geometries.

This work is part of work that is performed within the SESAME project of Horizon2020 research programme and is a continuation of research at our department.






10.09.2019 15:20 Thermal-hydraulics

Thermal-hydraulics – 203

Safety-related studies on heavy-liquid metal technology for advanced reactors in Europe

Walter Tromm

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

walter.tromm@kit.edu

 

As part of the European Sustainable Nuclear Industrial Initiative (ESNII), several innovative reactor concepts are being developed, with a roadmap for deployment of demonstrators by 2025. The Karlsruhe Institute of Technology (KIT) participates in the safety assessment of such concepts in the framework of the research program Nuclear Waste Management, Safety and Radiation Research (NUSAFE) and in close collaboration with European partners, e.g. in projects supported by Horizon 2020.

This work focuses on experimental investigations on heavy-liquid metal technology performed recently at KIT, relevant for the safety assessment of the MYRRHA and ALFRED reactors, planned for construction in Belgium and Romania, respectively. The experimental facilities available at the heavy-liquid metal laboratories of KIT are suitable for prototypical tests using molten lead or lead alloys at operating conditions (temperature, flow and power) representative of those expected in the reactors.

An overview of selected recent results is presented for two main disciplines, namely thermal-hydraulics and materials sciences. In the first category, heat transfer tests have been performed on rod bundles representative of the core fuel assemblies, with particular attention to the spacer geometry. Both nominal and non-nominal conditions, as required for licensing. The second category covers studies on the compatibility of structural materials in a liquid-metal environment at elevated temperatures. These include long-term corrosion tests with focus on the chemistry of dissolved oxygen as a protection mechanism for steel. Furthermore, the performance of surface-modification techniques developed at KIT for the in-situ formation of protective layers for are described. Moreover, an outlook on the research topics to be analyzed in the next years is presented.






11.09.2019 09:10 Materials, Integrity and Plant Life Management

Materials, Integrity and Plant Life Management – 308

oriFracture mechanics analyses of reactor pressure vessel under pressurized thermal shock with CFD and TRACE input loads

Diego F. Mora1, Oriol Costa Garrido2, Markus Niffenegger3, Roman Mukin4

1Swiss Nuclear Society Paul Scherrer Institute, Forschungsstrasse 111, CH-5232 VILLINGEN-PSI, Switzerland

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

3Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Reaktorstrasse, CH-5232 Villigen, Switzerland

4Paul Scherrer Institut, Villigen, CH-5232 Villigen PSI, Switzerland

diego.mora@psi.ch

 

Integrity assessments of a reactor pressure vessel (RPV) subjected to pressurized thermal shock (PTS) follow a multi-step simulation approach, where the thermo-hydraulic, thermo-mechanical and fracture mechanics simulations are solved sequentially, is presented and explained. Its main advantage is the modularity in which the different methods are assembled. One important aspect of this type of analysis is the resolution and qualification of the thermo-hydraulic analysis as it can affect the stress intensity factor (SIF) calculated in the fracture mechanics step. Since extreme thermal gradients in the structural components are expected during PTS, the fluid temperature must be reliably assessed to predict the loads upon the RPV. In case of non-uniform cooling down, a sophisticated way to calculate the complex flow field is needed. The use of computational fluid dynamics (CFD) calculations allows us to take into account local phenomena such as the cooling plume among others. This was done in previous studies carried out at the Paul Scherrer Institute for hypothetical small, medium and large break loss of coolant accidents (SB-, MB-, LBLOCA, respectively) in the hot leg of the two-loop pressurized water reactor (PWR). This gives refined results in terms of thermo-hydraulic key parameters such as temperature, heat transfer coefficient (HTC) and pressure. However, a disadvantage of these CFD analyses is that they are very time consuming and need large amounts of computational resources.
In order to reduce the computational cost, an alternative approach is the use of system codes such as TRACE. System codes have shown a noticeable progress in their capabilities and have reached an acceptable level of maturity. Furthermore, system codes have been qualified to analyze large and small break LOCA in conventional PWR. They solve the governing equations based on assumptions that disregard the fluid dynamics complexity. Initially, they were applied for the design of the engineered safety systems, but the development and elaboration of accident management procedures in the probabilistic safety analysis lead these codes to the best-estimate analysis. In this context, “best-estimate” means an accident simulation as realistic as possible. Therefore, in the best-estimate system codes, many conservative assumptions are replaced by the best-estimate approach for more realistic predictions. In this work, a three dimensional TRACE model is presented for the PWR. The cross verification of TRACE with CFD comparing key thermo-hydraulic parameter has shown that these are in good agreement with each other for high flow rates. Thus, all these facts allow us to use TRACE instead of CFD calculation in the thermo-hydraulic step of the multi-step simulation approach.
The output of TRACE simulation (i.e. temperature, HTC, pressure) is used as input in the subsequent thermo-mechanical and fracture mechanics simulations. The main objective is to determine the SIF of postulated cracks in the RPV wall and to compare them with those obtained using the input from CFD. The SIF calculation is performed in the framework of the extended finite element method in ABAQUS. This contribution presents three loss of coolant accidents of relevance for structural integrity of the RPV and the corresponding integrity assessments based on the thermo-hydraulic simulations using TRACE and CFD. The resulting SIF using both input shows good agreement. Therefore, it can be concluded that the implemented TRACE model is a good candidate for the integrity assessment of the PWR under consideration. The main difference is the behavior of the cooling plume in each case, which affects both stresses and stress intensity factors. The influence of the oscillating cooling plume in the integrity of the RPV is discussed as well as the advantages and disadvantages of each method.






11.09.2019 09:30 Materials, Integrity and Plant Life Management

Materials, Integrity and Plant Life Management – 302

EXPERIMENTAL STUDY OF PRIMARY RADIATION DEFECTS IN W AND Fe

Olga Ogorodnikova1, Mitja Majerle2, Volodymyr Gann3, Jakub Cizek4, Petr Hruska4, Stanislav Simakov5, Milan Stefanik2, Václav Zach2

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

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

3National Science Center ”Kharkov Institute of Physics and Technology”, 1, Academicheskaya str., Kharkov 61108, Ukraine

4Charles University in Prague Faculty of Mathematics and Physics, Prague, Czech Republic

5Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany

olga@plasma.mephi.ru

 

In future thermonuclear and advanced fission reactors, materials must withstand irradiation of high-energy neutrons and high-energy protons in space application. To predict the behavior of functional materials under irradiation, it is important to understand primary defect formation. Among the different experimental techniques available for the studies of defects in solids, such as electrical resistivity measurements, transmission electron microscopy and x-ray diffraction, only positron annihilation spectroscopy (PALS) allows us to resolve very small defects (> 0.1 nm) with very low concentrations (> 1 appm) and provides information on the size distribution of defects. The size of defect clusters plays a significant role in the long-term damage evolution and in the fuel accumulation. In this work, we successfully detected the radiation-induced defects in tungsten (W) and iron (Fe) irradiated with neutrons (n) with continuous spectrum up to 35 MeV and 22.5 MeV protons (p) in the dose range between 0.4×1020 and 2×1020 particles/m2. At room temperature irradiation, the average size of neutron irradiation-induced vacancy clusters is about 3 in W and 4 in Fe at lowest dose of 0.4×1020 n/m2 and increases with increasing the irradiation dose but more smoothly in Fe compared to W. At the same fluence, p-irradiation leads to a higher density of radiation-induced defects in W and Fe compared to n-irradiation (by factors of about ten and five, respectively). Preferentially mono-vacancies are formed in W during p-irradiation. On the other hand, small vacancy clusters consisting on average of 5 vacancies were observed in p-irradiated Fe. The density of radiation-induced vacancies in W is higher than that in Fe. One reason for this is the enhanced recombination of radiation-induced defects due to migration of vacancies in Fe at room temperature. To separate the recovery of radiation-induced defects by thermal annealing and by annealing of the hot recoils in cascade of collisions (athermal recombination, arc-dpa model), the irradiation has been performed at cryogenic temperature. This experiment indicated that the athermal recombination can be the dominant factor of the recovery of primary radiation-induced defects in Fe but the experimental data on radiation defects in W are well above the arc-dpa prediction indicating that the arc-dpa model needs further development. The effects of initial structural defects on primary radiation-induced defect formation have been also studied. For example, in the cases of pre-existing dislocations larger vacancy clusters after irradiation have been measured. The effects of initial dislocations and grain boundaries are discussed.






11.09.2019 09:50 Materials, Integrity and Plant Life Management

Materials, Integrity and Plant Life Management – 303

Probabilistic fracture mechanics analyses for structural integrity assessment of nuclear components

Oriol Costa Garrido1, Markus Niffenegger2, Diego F. Mora3, Roman Mukin4

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

2Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Reaktorstrasse, CH-5232 Villigen, Switzerland

3Swiss Nuclear Society, Paul Scherrer Institute, Forschungsstrasse 111, CH-5232 VILLINGEN-PSI, Switzerland

4Paul Scherrer Institut, Villigen, CH-5232 Villigen PSI, Switzerland

oriol.costa@ijs.si

 

Structural integrity assessments of nuclear components with crack-like imperfections are carried out in European countries and Switzerland following deterministic fracture mechanics (DFM) analyses. Due to their unique (deterministic) output, DFM analyses must employ conservative or best-estimate input data and assumptions. This could force the premature shutdown of aged nuclear power plants near the end of their operational lifetime. Supplementing the deterministic calculations, probabilistic fracture mechanics (PFM) analyses are also accepted in the USA. In this type of analyses, the uncertainties in the inputs, such as material properties and defects, are taken into account by their probability distributions. In this way, the interplay of the involved complex phenomena is considered in the integrity assessment. However, PFM computer codes are currently under development and their probabilistic outcomes being interpreted at a regulatory level.

With the aim to investigate the use of PFM analyses in nuclear applications, the Swiss national research project PROBAB (Probabilistic Component Integrity Analyses) was carried out at the Paul Scherrer Institut during 2016-2018. This paper summarizes the work performed during the last phase of the project, divided in two topics.

An integrity assessment of a reference reactor pressure vessel (RPV) under pressurized thermal shock (PTS) is performed assuming several loss of coolant accidents (LOCA), due to breaks of different sizes in the hot leg, as the initiating PTS events. The thermal-hydraulic transients following the breaks are studied with the system code TRACE and a three-dimensional model of the RPV downcomer and lower plenum. The fluid pressure, temperature and heat transfer coefficient output by TRACE are used as inputs in the PFM computer code FAVOR and in the finite element solver ABAQUS. The extensive TH screening analysis shows that a large break LOCA generally leads to severe PTS conditions. However, the results variability due to factors such as warm pre-stress effects (WPS) confirms the benefit of the PFM approach, the outcomes of which include the conditional probabilities of crack (growth) initiation and RPV failure.

In the framework of leak before break (LBB) evaluation, a dissimilar metal weld (DMW) joint containing inconel alloy 82/182 under active damage mechanism – stress corrosion cracking (SCC) and weld residual stresses (WRS) is analyzed with the PRO-LOCA computer code. In this study, plant in-service inspections and leak detection systems are also considered. The outcomes include the time-dependent probabilities of through-wall-crack (leakage) and pipe rupture, allowing for a probabilistic description of the LBB evaluation. The results indicate that SCC damage produces fast rupture of the piping soon after leakage initiates. However, the assumed leak detection of 10 gpm strongly decreases the rupture probability, possibly allowing the successful LBB case.






11.09.2019 10:50 Severe Accidents

Severe Accidents – 401

Sensitivity study on initial conditions of sodium vapour explosion in FARO-TERMOS T2 like conditions

Mitja Uršič, Matjaž Leskovar

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

mitja.ursic@ijs.si

 

Innovative sodium cooled fast reactors are considered as one of the future nuclear power plant reactor technologies. In the frame of safety studies, the risk for the environment in the case of a vapour explosion must be estimated. A vapour explosion can occur during the core melt accident when the rapid and intense heat transfer follows the interaction between the molten material and the coolant.

The comprehensive fuel-coolant interaction computer codes could be considered as a relevant tool for the safety studies related to the issue of the vapour explosions. Currently, the applicability of the MC3D code (IRSN, France) to simulate the fuel-sodium interaction is under examination.

The purpose of this paper is to discuss the MC3D applicability to simulate in the FARO-TERMOS T2 (JRC, Ispra, Italy) experiment like conditions. Namely, due to the large melt mass, the performed FARO-TERMOS experiments are considered closest to the real situation. Moreover, the occurrence of spontaneous vapour explosions was reported. The objective is to perform a sensitivity study to get insights into the initial conditions at the time of the vapour explosion triggering.






11.09.2019 11:10 Severe Accidents

Severe Accidents – 402

A Separate Effect Study of the Influence of Molybdenum on Iodine and Caesium Transport in the Primary Circuit in Nuclear Severe Accident Conditions

Melany Gouello1, Jouni Hokkinen1, 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

 

In case of a severe accident on a Light Water Reactor (LWR), iodine may be released into the environment, impacting significantly the source term. The understanding of the iodine release from the damaged reactor core and its transport in the different parts of the reactor up to the reactor containment, is thus a major issue. The analysis of the Phébus-FP tests brought the hypothesis that iodine keeps a gaseous form up to the containment due to some processes that limit the formation of caesium iodide in the reactor coolant system (RCS). The formation of caesium iodide, which is assumed to be the dominant form of iodine in the RCS, would be notably limited due to the presence of other elements reacting with caesium.
An experimental study has been launched at VTT investigating the behavior of iodine on primary circuit surfaces during a severe nuclear accident. Caesium iodide and molybdenum trioxide were used as a nonradioactive precursor materials in order to highlight the effects of carrier gas composition and oxygen partial pressure on the chemistry and transport. Aerosols and gaseous species released from the reaction crucible 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) as well as with Electric Low Pressure Impactor (ELPI).
The study investigated, first, the influence of molybdenum presence on the caesium iodide behavior at 700 °C under two atmospheres: Ar/H2O and Ar/Air. Experiments showed that the transport of gaseous iodine was enhanced by the presence of molybdenum, explained by the formation of a caesium molybdates in the crucible. This suggests that a reaction between condensed caesium iodide and molybdenum is possible at the surface of primary circuit in these conditions. In addition, the oxygen partial pressure prevailing in the studied conditions was determined as an influential parameter in the reaction. The effect of molybdenum on the release of gaseous iodine from CsI-MoO3 precursor mixtures with different Mo/Cs molar ratios and atmosphere compositions was investigated. Effect of the initial amount of powder was also studied.






11.09.2019 11:30 Severe Accidents

Severe Accidents – 403

Modelling of Severe Beyond-Design-Basis-Accident at Spent Fuel Pool of AES-2006 Taking into Account the Fuel Elements Oxidation Features in Steam-Air Atmosphere

Alexander D. Vasiliev1, Yurii Zvonarev2, Valerii Merkulov2

1Nuclear Safety Institute of Russian Academy of Sciences , 52, B. Tulskaya, 115191 Moscow, Russian Federation

2National Research Center «Kurchatov Institute», 1, Akademika Kurchatova pl., Moscow, 123182, Russian Federation

vasil@ibrae.ac.ru

 

In the course of postulated severe beyond-design-basis-accident (BDBA) with loss of coolant in a spent fuel pool (SFP) of NPP AES-2006 a large amount of hydrogen can be released due to chemical reactions of zirconium and steel with containment atmosphere consisted of steam, oxygen and nitrogen. A highly explosive mixture of hydrogen and oxygen may threaten the integrity of containment. Besides, the heat-up of SFP fuel assemblies (FAs) may result in their destruction, melting and release of radioactive fission products (FP) to the atmosphere of containment and, after loss of its integrity, to the environment.
This is why a comprehensive analytical and numerical analysis of basic processes occurring in the course of loss-of-coolant BDBA is needed. It is necessary to take into account that thermal hydraulic, thermo-mechanical, chemical and other relevant phenomena are very complicated and still far from full understanding.
Basically, a loss-of-coolant accident in SFP can be divided in four characteristic phases:
• heat-up of SFP water to boiling temperature;
• gradual water level drop to the top of heated part of FAs;
• severe accident phase with FAs heat-up, oxidation and hydrogen generation, their destruction and mass relocation to lower elevations;
• corium-concrete interaction, concrete destruction and hydrogen generation.
For adequate modeling of that accident with the use of severe accident code it is necessary to take into account specific features of physical-chemical processes in the course of SFP accident compared to the processes taking place at reactor in-vessel phase processes for modeling of which a majority of severe accident codes are intended. In particular, the zirconium-based claddings will occur in the atmosphere containing oxygen and nitrogen apart from steam, which is supported by two circumstances: 1) the oxygen density is higher than the steam one which makes easier for oxygen to reach FAs; 2) Zr oxidation reaction by oxygen has higher priority compared to steam reaction.
The Zr-based claddings high-temperature oxidation in steam-oxygen-nitrogen atmosphere behaves itself completely different (much more aggressively) compared to oxidation in pure steam medium. Not taking into account of this complicated oxidation kinetics will result in considerable (by several tens of percents and probably by several times of magnitude) underestimation of hydrogen generation rate and integral production.
In this paper, the numerical modeling of severe beyond-design-basis-accident with loss of coolant in a SFP of NPP AES-2006 is conducted using the SOCRAT/V2 code. The results of basic output calculation parameters for several characteristic scenarios are presented. It is shown that even using of standard steam oxidation kinetics results in very considerable hydrogen release (several tons, depending on SFP loading and accident scenario). Taking into account of oxidation features in gas atmosphere containing oxygen, steam and nitrogen, may result in additional considerable increase of both integral hydrogen generation (by several tens of percent) and hydrogen generation rate (up to several times).






11.09.2019 11:50 Severe Accidents

Severe Accidents – 404

Uncertainty Analysis of the Hydrogen Production in PHEBUS FPT-1 Experiment

Piotr Mazgaj, Piotr Darnowski, Grzegorz Niewiński

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

piotr.darnowski@pw.edu.pl

 

The paper describes the MELCOR simulations of the core degradation during the bundle phase of the PHEBUS FPT-1 experiment. The uncertainty analysis was performed with particular emphasis put on hydrogen production. The FPT-1 model was developed with MELCOR 2.2 and publicly available data. The main results, the temperatures of fuel rods and the shroud, the oxidation and resulting hydrogen production, were discussed and successfully verified with literature results. Finally, the assessment of the new MELCOR 2.2 was performed for the prediction of the main degradation phenomena together with uncertainty analysis based on the GRS method.






11.09.2019 12:10 Severe Accidents

Severe Accidents – 405

Features and Validation of the New Transport Module in ATHLET-CD

Livia Tiborcz, Thorsten Hollands

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

livia.tiborcz@grs.de

 

Deterministic safety analyses are of great importance to assess existing as well as future nuclear power plants. The thermal-hydraulic system code ATHLET-CD, as part of the German code package AC2, is developed by GRS for the comprehensive analyses of severe accidents in nuclear power plants. The module responsible for the transport and deposition of fission product in the reactor cooling system has been greatly changed between the versions 31A and 32. The old module, SOPHAEROS, has been replaced by a new one SAFT (Simulation of Aerosol and Fission Product Transport). SAFT is based on SOPHAEROS available in ASTEC V2.0, which has numerous improvements compared to the previously available version in ATHLET-CD. Furthermore, thanks to a new implementation method it can simulate branching, which enables a more realistic simulation of fission products and aerosol behavior especially during plant analyses. After a short description of the new transport module a validation calculation based on Phébus FPT3 is presented. The results by the old and new versions are compared and the improvements of the new SAFT module are discussed.






10.09.2019 09:10 Research Reactors

Research Reactors – 501

Design and safety of the PGNAA facility at the TRIGA Research Reactor of the University of Pavia – LENA

Andrea Salvini1, Daniele Alloni1, Massimiliano Clemenza2

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

2INFN, Largo Enrico Fermi, 2, I-50125 Firenze, Italy

andrea.salvini@unipv.it

 

A Prompt Gamma Neutron Activation Analysis (PGNAA) facility has been recently implemented at the TRIGA MARKII nuclear research reactor of the LENA Laboratory (University of Pavia, Italy) developing a new neutron beam with a proper irradiation room and shielding from one of the available horizontal channels of the reactor. This is the result of a two-year work within the framework of the INFN-CHNet project (National Institute of Nuclear Physics – Cultural Heritage Network). This paper reports the main steps of the design and the main characteristics of the facility, particularly focussing on the safety aspects of the project.






10.09.2019 09:30 Research Reactors

Research Reactors – 502

Experimental validation of RAPID for calculation of flux redistribution factors due to control rods insertion in TRIGA reactors

Valerio Mascolino1, Alireza Haghighat1, Luka Snoj2

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

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

val@vt.edu

 

The RAPID Code System, based on the Multi-stage Response-function Transport (MRT) methodology, is a software for real-time neutronics analysis of nuclear systems. RAPID is capable of calculating criticality fission neutron source distributions (both prompt and delayed), criticality eigenvalue, subcritical multiplication factor, kinetic parameters such as beta effective and Alpha-Rossi, detector responses, and time-dependent neutron source distributions.
For calculating source distributions and parameters such as k-effective, Alpha-Rossi, and effective delayed neutrons fraction, RAPID utilizes the Fission Matrix (FM) method. The method relies on accurate pre-calculation of coupling coefficients among the various fissionable regions within the nuclear system. The coefficients are calculated within ranges of relevant parameters (e.g., fuel burnup for spent nuclear fuel systems or control rods position for TRIGA reactors) to generate an FM database. Then, the coefficients are combined and interpolated in real-time to solve a linear system of equation in real-time for any system configuration.
Recently, as part of an ongoing collaboration between Virginia Tech (VT) and the Jožef Stefan Institute (JSI), we have developed a new FM-based algorithm for simulation of movement of Control Rods (CRs) within nuclear reactors. The methodology, referred to as “FM-Control Rods deltas” (FM-CRd), allows to significantly reduce the pre-calculations required to compile a full FM database to account for the movement of control rods within the reactor. Each control rod effect is accounted within the CRd’s independently. The CRd’s are then linearly combined with an “all-rods-out” (ARO) FM before solving the FM equation. Results from the steady-state validation of the methodology using JSI’s TRIGA in its Core 133 configuration (as per the criticality benchmark included in the ICSBEP handbook) will be shown in the paper.
The goal of the joint VT-JSI project is to be able to utilize the RAPID Code System for live calculation of the flux redistribution factors required for evaluation of control rod worth measurements. To this end, we have conducted an experimental campaign on February 2019. The experimental campaign consisted in performing the rod-swap and rod-insertion experiments while collecting detector data from two sets of different in-core fission chambers. The fission chambers were positioned at different locations, radially and axially, within the core. Experimental data for the two sets of experiments will be presented.
While the rod-swap experiment can be assumed to go through a series of quasi steady-state steps, the rod-insertion requires to take into account the variation in time of the delayed neutrons within the system. Modeling this phenomenon requires Time-dependent FM’s (TFMs) to perform kinetic simulations. As mentioned, RAPID is capable of calculating time-dependent neutron source distributions. The time-dependent module of RAPID is referred to as tRAPID.
This work will compare RAPID results to measurements and Monte Carlo Serpent predictions for both rod-in and rod-swap experiments.






10.09.2019 09:50 Research Reactors

Research Reactors – 503

SENSITIVITY AND UNCERTAINTY ANALYSIS OF THE TRIGA MARK II REACTOR USING FIRST AND SECOND ORDER PERTURBATION THEORY

Luis Daniel Celeita Perez1, Christian Castagna2, Stefano Lorenzi3, Antonio Cammi2

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

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

3Multiple organizations possible, Unknown, Unknown, Slovenia

christian.castagna@polimi.it

 

During the last years, perturbation techniques have been widely implemented for the analysis of the sensitivity and propagation of uncertainty in simulations of physical systems, such as nuclear reactors, due to changes in control or external parameters. The use of first and second-order perturbation theory allows the generation of perturbation coefficients employing a weight function used to define the response of interest and the solution of the system of differential adjoint equations. If slightly variations of the operational parameters as the coolant inlet temperature or the reactivity occurs, the first order perturbation theory is sufficient to obtain satisfactory results, especially if the focus is on the steady-state values at the end of the transient. In particular, the adjoint perturbation theory has different performances according to the section of the transient is under investigation. If the interest involves all the duration of the transient, the second term correction is needed to obtain acceptable results. The latter are also affected by the selection of the variables subject to variation of the control parameter and to which the perturbation theory is applied.
This paper discusses the application of the first and second order adjoint perturbation theory to the analysis of the sensitivity and uncertainty propagation of the Pavia TRIGA Mark II reactor. A simplified model [1] is used to describe the evolution of the neutron flux, the fuel and coolant temperature and their feedback on the reactivity of the system. The results show that second-order corrections are able to reproduce the response system with a slight difference with respect to the direct solution of differential equations. Errors lower than 0.1% (compared with the direct solution) are found with variations of up to 10% of the control parameter. The approach can be easily extended to take into account variations of reactivity by effect of the accumulation of neutronic poisons in long transients [2] or eventually attaching the entire heat transfer problem, taking into account the heat exchangers subsystems and variations in the water supply.

[1] Cammi, A., Ponciroli, R., Borio di Tigliole, A., & Magrotti, G. (2013). A zero dimensional model for simulation of TRIGA Mark II dynamic. Progress in Nuclear Energy, 68, 43-54. Retrieved 2019, from www.elsevier.com/locate/pnucene
[2] Introini , C., Cammi, A., Lorenzi , S., & Magrotti, G. (2019). An improved zero-dimensional model for simulation of TRIGA Mark II. Progress in Nuclear Energy, 85-96. Retrieved from www.elsevier.com/locate/pnucene






10.09.2019 10:10 Research Reactors

Research Reactors – 504

Analysis of the contribution of the gamma flux for the incore neutron flux measurement in MTR

Christophe Destouches1, Vladimir Radulović2

1Commissariat a l’Energie Atomique – Centre d’Etudes de Cadarache DRN/DER, Bat 238, 13108 St Paul Lez Durance Cédex, France

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

christophe.destouches@cea.fr

 

Neutron flux measurements are always performed in (?,n) mixed field especially in the core of a MTR where the both fluxes are generally of the same order of magnitude. Therefore, the contribution of the gamma particle has to be taken into account in the sensor design or in measurement methods for minimizing their impact or correcting it. The latter is necessary to be evaluated for an accurate measurement of the neutron flux in perturbed areas where the ratio of the two kind of particle fluxes are strongly variating in time or in space. It is encountered for example in the transition zone between the core and the reflector or in some low power reactor dynamic states. In addition, gamma energy spectra and level are usually evaluated with a coarse uncertainty by modelling tools, even if some recent upgrades have been done in this domain in terms of modelling and measurements.
After a review of the different reference techniques applied to the measurement neutron flux (dosimetry, fission chamber, SPND), their performances and limitations in terms of gamma sensibility are analysed. Then, based in the expertise of the CEA and JSI teams, propositions of combining of existing measurement methods and modelling results are proposed to optimize the measurement process. In addition, propositions for improvements of some measurement methods are given based on the impact on the gain on the accuracy of the measurement process, in particular for comparison with calculations.






10.09.2019 10:50 Reactor Physics

Reactor Physics – 601

SuperFINIX — A flexible-fidelity core level fuel behavior solver for multi-physics applications

Ville Valtavirta

VTT, Tietotie 3, FI-02150 Espoo, Finland

ville.valtavirta@vtt.fi

 

VTT Technical Research Centre of Finland Ltd (VTT) is currently re-building its computational reactor analysis framework. The development of this new framework, Kraken has recently started and the framework builds on in-house developed solvers such as the neutronics solvers Serpent (Monte Carlo continuous energy) and Ants (multi-group nodal) as well as the fuel behaviour solver FINIX (traditional 1.5-dimensional rod representation). The separate solvers communicate with a central multi-physics driver that transfers field data and controls the general solution flow. The individual solvers are coupled in a modular fashion, meaning that the solver for a specific task can be switched to one with a higher fidelity or a lower running time if needed.

While the solvers for the neutronics and for thermal hydraulics consider the whole reactor core as their solution domain, the fuel behaviour solver FINIX only models a single fuel rod. As current light water reactor cores contain tens of thousands of individual fuel rods, a separate tool that distributes the task of solving the core level fuel behaviour to tens of thousands of FINIX instances is needed. An additional consideration in the context of the Kraken framework is the different fidelities supported by the neutronics solvers. While Serpent can tally the pin-power distribution using a detailed axial-radial meshing inside each individual fuel rod if needed, the nodal code Ants is limited to a much coarser resolution.

SuperFINIX is a recently created wrapper code for multiple FINIX-instances that accepts the core level power distribution from the multi-physics driver, distributes it to FINIX-instances describing the fuel rods in the core, executes the FINIX solvers, collects the fuel behaviour result and passes it back to the multi-physics driver. While this simple task is important in itself, SuperFINIX is tailor made for variable-fidelity multi-physics coupling so that the same SuperFINIX instance can accept power distributions and provide fuel temperature distributions at multiple different levels of fidelity. In the common case, each individual fuel rod is initialized individually and any coarser level power distributions that are provided (e.g. assembly or sub-assembly level power distributions) are passed to the individual fuel rod instances. The fuel behaviour solution is solved for each rod separately and can be provided as axial-radial distributions for each individual rod or homogenized to some lower fidelity, e.g. pin-wise temperatures without radial dependence or assembly-wise (or sub-assembly-wise) temperatures).

This paper will describe SuperFINIX including its currently available temperature homogenization procedures and the parallel execution of the individual rod-level FINIX solvers. The use of SuperFINIX will be demonstrated in a minicore case, coupled to the Serpent Monte Carlo code (and potentially to the Kharon channel level thermal hydraulics solver). Comparisons will be made between coupled solutions obtained using different fidelities for the fuel behaviour feedback (different fidelities of fuel temperature) and using different approaches for fuel temperature homogenization.






10.09.2019 11:10 Reactor Physics

Reactor Physics – 602

Evaluating the effect of decay and fission yield data uncertainty on spent nuclear fuel source term using Serpent 2

Antti Rintala

VTT, Tietotie 3, FI-02150 Espoo, Finland

antti.rintala@vtt.fi

 

Knowledge of spent nuclear fuel (SNF) source term (decay heat, reactivity, nuclide inventory and other relevant properties of SNF) is essential in the safe handling and final disposal of SNF. For example, decay heat power determines how densely the fuel canisters can be packed in the final disposal tunnels. Computational characterization of SNF involves numerous uncertainty sources, one of which is the uncertainty in nuclear data. Different nuclear data libraries are known to yield significant differences in nuclide inventory calculations. How these uncertainties affect each component of the source term is not generally known. This work aims to identify the major uncertainty components in decay and fission yield data in relation to the SNF source term.

The uncertainty propagation from decay and fission yield data uncertainties into the uncertainties of the source term will be conducted using a sampling based technique and an extended version of the Monte Carlo particle transport code Serpent 2. Normally, Serpent 2 only uses the tabulated values for the fission yield and radioactive decay data from ENDF-6 format data files. In the extended version, the nominal values are perturbed based on random sampling using the uncertainty data present in the ENDF-6 format files. Producing the SNF source term several times using different randomly perturbed nuclear data allows the propagation of the input uncertainties to all output quantities included in the SNF source term. Sub-components of the resulting output uncertainty can be identified by limiting the perturbation to a specific class of data, e.g. by only perturbing the fission yield data of U-235 and Pu-239.

The analysis will investigate the decay and fission yield data related uncertainties in the SNF source term of a TVEL second-generation type VVER-440 fixed assembly with an average enrichment of 4.37 % U-235 and six gadolinium-doped fuel rods with 3.35% Gd2O3. Other nuclear data related uncertainties are ignored in this study, and only fixed, nominal depletion conditions are considered.






10.09.2019 11:50 Reactor Physics

Reactor Physics – 604

Neutron Spectrum Unfolding Exercise REAL-2019

Andrej Trkov1, Ingrid Vavtar2

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

a.trkov@iaea.org

 

Neutron spectrum is an important characteristic of a neutron field because activation, radiation damage and health hazards depend on the energy of the neutron. Neutrons are neutral particles and cannot be detected by their mere presence – they are detected when they undergo a reaction. After interaction the target nuclei are often radioactive, emitting characteristic radiation that can be used to uniquely identify the target nucleus. This property is used as one of the methods to detect neutrons after relatively long exposures using a variety of monitor reactions. Since the reaction rate depends on the cross sections of the monitor, which are energy-dependent, this property is utilized to provide information on the neutron spectrum by the so-called unfolding procedure.

In 1984 an exercise was conducted to compare the methods and codes for neutron spectrum unfolding. Given a set of measured reaction rates and a trial spectrum of selected irradiation facilities, participants of the exercise analysed the data and compared the obtained spectra resulting from the unfolding process with their codes. The exercise was repeated in 1988 using refined techniques. Since then the computational capabilities have increased enormously, allowing more accurate computational simulation of the experimental set-up to obtain the prior neutron spectrum shape. Also, many additional experimental measurements of reaction rates became available. Participants of the IAEA Co-ordinated research project on the “Testing and Improving the International Reactor Dosimetry and Fusion File (IRDFF)” at the Meeting in 2017 recommended that the exercise should be repeated using new dosimetry library IRDFF-II that is being tested prior to general release.

Suitable experimental facilities were identified, which are well characterized and on which suitable measurements of reaction rates were made with dosimetry materials included in the IRDFF-II library. The data were checked for completeness and are being uploaded to the web server at the IAEA for the use of the participants of the exercise.
Detailed description of the project and its scope will be presented in more detail. Eligible participants should be proficient in the use of analysing activation data of monitor reactions with an unfolding code of their choice. The results and the methods will be compared, providing guidance on the optimal procedures for neutron spectrum determination.






10.09.2019 12:10 Reactor Physics

Reactor Physics – 605

Implementation of Neutron Trap Passive System in GFR 2400 Fast Reactor Design

Filip Osuský, Branislav Vrban, Ján Haščík

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

filip.osusky@stuba.sk

 

The paper investigates implementation of neutron trap passive system within a gas-cooled fast reactor design, referred as GFR 2400. This concept was identified by Generation IV International Forum as prospective fast reactor design, however technical feasibility of this concept is questionable. Difficulties may arise during unprotected transient overpower and unprotected loss of flow scenarios. Therefore, the implementation of the neutron trap passive system may mitigate the consequences of these scenarios. The assumed passive actuation mechanism of the neutron trap is a Currie point latch with Fe-Ti-O ferromagnetic material, which Currie point temperature is 850 °C.
The NESTLE code is used for simulation of transient coupled calculation. The NESTLE code solves multigroup neutron diffusion equation by finite difference method or by nodal expansion method and the solution of this equation is internally coupled with thermal-hydraulic sub-channel code. The nodal expansion method invalidates symmetry of the material matrixes used in numerical solution of the diffusion equation. Although, this correction of nodal expansion method has been found relatively small for the systems such as BWR and PWR, it is not clear if this method is also suitable for fast reactor systems. Therefore, the finite difference method is used for the transient simulation of GFR 2400. The SCALE code system is used for processing the homogenized multigroup macroscopic cross-section library that is suitable for the NESTLE code. To provide correct neutron spectrum for weighting of macroscopic cross-section library, the investigated assembly is surrounded by the homogenous representation of the GFR 2400 core. Developed standard genetic algorithm is used for the optimization of nuclide composition of the surrounding homogenous mixture to achieve similar reaction rates on particular nuclides in the investigated assembly in comparison with GFR 2400 full core model.
Two transient scenarios are investigated, where the first assumes rapid withdrawal of one control rod assembly during normal operation and the response of the neutron trap passive system is observed. The second assumes continuous shutdown of one primary circuit blower from the total three blowers that are incorporated in GFR 2400 design. In this case also the response of the passive system is investigated. Changes in temperature distributions are described and visualized in the paper.






12.09.2019 09:10 Nuclear Fusion

Nuclear Fusion – 701

Validation and Use of Coupling SUSD3D with Denovo for Complex Sensitivity/Uncertainty Analysis

Bor Kos, Ivo Aleksander Kodeli

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

bor.kos@ijs.si

 

Bor Kos, Ivan Kodeli and JET contributors*
*See the author list of X. Litaudon et al 2017 Nucl. Fusion 57 102001

ABSTRACT
Nuclear data uncertainty propagation through the particle transport calculation is a vital part of the computational uncertainty determination. Uncertainties due to nuclear data usually exceed modelling (discretization) and/or statistical uncertainties connected with the use of deterministic and stochastic transport methods. Exact knowledge of these uncertainties is not only needed for benchmarking purposes, nuclear data validation and adjustment but also to determine and eventually reduce safety margins in reactor and shielding designs.
Different methods can be used to propagate nuclear data uncertainties, such as random sampling and perturbation. Random samples of nuclear data libraries created in accordance to the covariance matrices are used in a series of transport calculations. The distribution of this series of results gives us the final uncertainty. This approach, also known as Total Monte Carlo, is time consuming or even impossible for deep shielding cases where one such (a single) calculation(s) (can) takes a few hours (or even days), and several hundreds of calculations need to be performed to gain statistically relevant results.
Alternatively codes based on perturbation theory can be used where only one forward and one adjoint calculation need to be performed to gain sensitivity and uncertainty information for every reaction included in the nuclear data libraries. A perturbation based Sensitivity/Uncertainty (S/U) code SUSD3D [1] was developed at NEA Data Bank and JSI and routinely used for complex fission, fusion, medical and industrial applications. Recently the code was coupled with the Denovo transport solver. Denovo [2] is used as part of the ADVANTG [3] package which significantly reduces the user time needed for execution of deterministic calculations.
Several auxiliary programs have been written to aid the user when performing an S/U calculation with SUD3D. The resulting code package – SUD3DwD (SUSD3D with Denovo) – has been validated on the ASPIS Iron 88 benchmark experiment against results of SUSD3D coupled with the 2D transport code DORT, 3D transport code Partisn and against results calculated using a random sampling based code SANDY [4] developed at SCK•CEN.
SUSD3DwD was also used to perform the first preliminary nuclear data sensitivity and uncertainty analysis of the JET3-NEXP streaming benchmark experiment performed at the Joint European Torus.
ACKNOWLEDGMENT
This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training program 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.
REFERENCES
[1] Kodeli, I., Slavic, S., “SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross Section Sensitivity-Uncertainty Analysis“, Science and Technology of Nuclear Installations, Volume 2017, Article ID 1264736, (2017), https://doi.org/10.1155/2017/1264736.
[2] T. M. Evans, A. S. Stafford, R. N. Slaybaugh, and K. T. Clarno, Denovo: A New Three-Dimensional Parallel Discrete Ordinates Code in SCALE. Nuclear Technology, 171, 171-200 (2010).
[3] S.W. Mosher. (2015). ADVANTG-An Automated Variance Reduction Parameter Generator, ORNL/TM-2013/416 Rev. 1, Oak Ridge National laboratory.
[4] L. Fiorito, G. Žerovnik, A. Stankovskiy, G. Van den Eynde, P.E. Labeau, “Nuclear data uncertainty propagation to integral responses using SANDY,” In Annals of Nuclear Energy, 101, pp. 359-366 (2016).






12.09.2019 09:30 Nuclear Fusion

Nuclear Fusion – 702

Thermal Modelling of ITER First Wall Panels

Matic Brank1, Leon Kos1, Richard Pitts2, Gregor Simic1

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

2ITER Organization, Cadarache Centre, Building 519, 13108 St. Paul lez Durance, France

matic.brank@lecad.fs.uni-lj.si

 

The ITER tokamak is designed to produce fusion grade plasmas for durations of hundreds to thousands of seconds. These long durations require active cooling of the wall armour directly facing the plasma. Such actively cooled plasma-facing components (PFCs) have strict limits on the allowed stationary surface heat flux densities in order to avoid critical heat fluxes at interfaces with the cooling channels. These flux densities are lower for the beryllium (Be) main chamber PFCs than for the tungsten divertor targets and must be carefully monitored from the beginning of ITER operations, when the Be first wall panels (FWP), notably in the inboard equatorial region, will be used for plasma start-up. It is thus extremely important to provide assessments of likely heat loading and thermal transfer within the Be armour for incorporation into real time plasma control algorithms.
This article describes a new module in the SMITER magnetic field line following code framework [1] for finite element calculations of thermal transfer in FWPs subject to heat flux distributions on the front surface computed by SMITER. The heat fluxes are mapped onto a 3D model of the panel subcomponents, comprising, for each FWP, a series of poloidally stacked “fingers”, each of which is made up of a cooled substrate, to which are bonded individual Be tiles. The ITER FWPs come in two variants: Enhanced (EHF) and Normal (NHF) panels. Models for both are included in the SMITER thermal module.
Computation of static temperatures on the FWP surface is performed using the open source FEM package Elmer [2], which is embedded in the SMITER GUI environment. For selected EHF FWPs, where plasma-wall interactions are expected to be highest, the full Be tile castellation geometry is included in the CAD model, yielding a computationally heavy task for field line tracing and thermal transfer, but providing a much higher fidelity representation of the surface temperature distribution than obtained with the standard smooth surface profile used for most heat flux distribution estimates. The GEOM module, also part of SMITER GUI environment and having an extensive library for Python scripting, is used to prepare CAD models for the structures below the Be castellations. Meshing of the models is performed with SMESH, a further module embedded in the SMITER GUI.
A full tutorial from CAD modelling and meshing to temperature calculation will be present in this article together with the result of some benchmarking of the new framework against specific existing thermal calculations provided by the ITER Organization.






12.09.2019 09:50 Nuclear Fusion

Nuclear Fusion – 703

Influence of surface roughness on sputtering of Mo under keV D ions irradiation

Mitja Kelemen1, Primož Pelicon1, Sabina Markelj1, Espedito Vassallo2, David Dellasega3, Matteo Passoni3, Matteo Pedroni4, Thomas Schwarz-Selinger5, Antti Hakola6, Eurofusion Wp Pfc Contributors

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

2Istituto Superiore di Sanita, Viale Regina Elena, 299, I-00161 Rome, Italy

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

4Istituto per la Scienza e Tecnologia dei Plasmi, Via Roberto Cozzi, 53, 20125 Milano, Italy

5Max-Planck-Institut für Plasmaphysik (IPP), Boltzmannstr. 2, D-85748 Garching b. München, Germany

6VTT, Tietotie 3, FI-02150 Espoo, Finland

mitja.kelemen@ijs.si

 

An important issue in the development of thermonuclear fusion is the lifetime of the reactor walls. Irradiation by ions and neutrals will lead to erosion of the wall materials, contaminating the plasma and reducing the lifetime of plasma facing components. For perfectly smooth surfaces a theoretical model predicts sputter yields with great accuracy [1] showing a distinct angular dependence of the sputter yield [2]. However, when dealing with rough surfaces the experimental results deviate from model predictions and also the angle dependence is less distinct. Within the Eurofusion work package PFC a dedicated task was launched to quantitatively determine the influence of roughness on the sputter yield [3].
To address this issue erosion of molybdenum (Mo) thin films on graphite substrate by deuterium (D) ions was investigated. Such composition of samples was chosen on the basis that same studies ants to be carried out in ASDEX Upgrade. For this purpose we constructed an experimental setup where samples are exposed to D ions generated by an electron cyclotron resonance ion gun. The dominating ion species coming out from the ion gun are D3+ ions, which are accelerated by applying 3.1 kV voltage in this case and decelerated in front of the sample to 3.0 keV, therefore yielding 1 keV energy per D.
120 nm thick Mo films deposited by pulsed laser deposition on graphite substrate of different roughness were produced for this purpose. The surfaces roughness (Ra) ranged from mirror polished surfaces (Ra= 5 nm) to rough surface (Ra= 2-3 µm). Samples were exposed to 1.5×1023 D ions/m2 at different impact angles with respect to the surface (0°, 40°, 60° and 70°). The erosion was determined by measuring Mo areal density before and after the exposure by Rutherford Backscattering Spectroscopy (RBS). The RBS measurements were performed by 2.5 MeV 4He ion beam. In the presented contribution results of the effect of surface roughness on sputter yields and its angular dependence will be discussed in details as we observed clear angular dependence for Ra~5 nm which is smeared out at higher surfaces roughness.

[1] W. Eckstein Top. Appl. Phys. 110 (2007)
[2] M. Küstner et al. J. Nucl. Mater. 265 (1999)
[3] R. Arredondo et al. Nucl. Materials and Energy 18 (2019)






12.09.2019 10:30 Radiation and Environmental Protection

Radiation and Environmental Protection – 806

PERFORMING MEASUREMENTS IN FUKUSHIMA

Tamara Gregorčič, Michel Cindro, Samo Tomažič

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

tamara.gregorcic@gov.si

 

On March 11th 2011 a tsunami severely damaged Japan’s coastal areas, including the Fukushima Daiichi Nuclear Power Station. Substantial amounts of radioactive materials were released into the environment. In an event of a nuclear or radiological accident it is essential to promptly provide data on radioactivity in the environment, since the successful implementation of protective measures for the population depends on this data. Since 2013, the International Atomic Energy Agency (IAEA) holds International Emergency Response Workshops (RANET) in Fukushima. These five-day workshops aim at further strengthening nuclear and radiological preparedness and response capabilities of member states.
In August 2018, two employees from the SNSA`s Monitoring Section attended the RANET workshop. During the workshop, participants from eleven countries conducted radiation monitoring and environmental sampling and analysis. Participants measured the contamination level of the ground surface and conducted gamma spectrum analysis and vehicle-based monitoring – activities that would follow any nuclear or radiological incident or emergency. Results were compared amongst participants.
Workshop also presented a great opportunity to test SNSA`s recently upgraded software and mobile application called Radioactivity In The Environment (RVO) in the real circumstances with higher dose rates. The RVO system is an automatic system for gathering and reporting radioactivity data measured in the environment with the purpose of prompt detection and alarming in the case of elevated values in the environment in the event of an emergency. Additionally, the SNSA had developed the RVO mobile application to be used by the field teams to communicate real-time data to the decision makers and display it on the RVO web portal.
SNSA´s instrumentation setup consisted of Automess 6150AD and IDENTIFINDER R400, Bluetooth interface to phone and android/iPhone app. During an extensive two-day testing period in the vicinity of the Fukushima Daichi NPP the SNSA’s team faced some challenges like integration time vs. walking speed, looking for representative values or hot spots, waterproofing and interruptions of data transfer.
Bringing together so many experts from different countries in one place, the workshop helped the SNSA team learn how international teams can work together to provide assistance in a nuclear or radiological emergency situation. The workshop gave participants the opportunity to learn from each other and helped gain essential feedback about the RVO´s mobile application and ideas for further development.






12.09.2019 10:50 Radiation and Environmental Protection

Radiation and Environmental Protection – 802

Radiation transport capabilities in the Serpent 2 Monte Carlo code

Jaakko Leppänen

VTT Technical Research Centre of Finland Ltd., Kivimiehentie 3, 02044 VTT, Finland

jaakko.leppanen@vtt.fi

 

The Serpent Monte Carlo code was originally developed as a computational tool for various neutron transport problems encountered in reactor physics applications, but in recent years the scope has been broadened to new fields, including radiation transport and fusion neutronics. The development work has been focused on three topics: 1) Photon transport calculations involving source terms comprised activated materials; 2) Advanced CAD-based geometry types and 3) Weight-window based variance reduction methods with a built-in importance solver.

The photon physics model was originally implemented for the purpose of accurate heat deposition calculations in reactor analysis, but the capability is also applicable to general radiation transport applications. The radioactive decay source mode combines material compositions obtained from a separate burnup or activation calculation with ENDF decay spectra, and forms the source terms automatically without additional user input.

The CAD-based geometry type is based on the STL data format, which is widely used for 3D printing and therefore supported by virtually every CAD tool. Geometries comprised of STL solids are handled as separate universes, which can be combined with conventional CSG and other available geometry types. Apart from exporting the native CAD format into STL, the procedure involves no format conversions.

Variance reduction in Serpent is based on a conventional mesh-based weight-window technique. The weight-window mesh can be read from MCNP WWINP format files, or produced using an built-in solver based on the response matrix method. The routine is capable of calculating importances with respect to one or multiple responses, and an iterative global variance reduction scheme can be used to gradually expand the mesh throughout the whole geometry. The supported mesh types include conventional Cartesian, cylindrical and hexagonal mesh, but also unevenly-spaced and adaptive octree mesh types.

The purpose of this paper is to provide a general overview and practical examples of the methodology used in Serpent for radiation transport calculations, including its advantages, major limitations and future prospects.






12.09.2019 11:10 Radiation and Environmental Protection

Radiation and Environmental Protection – 803

Evaluation of the NEK SFDS cask model using hybrid shielding methodology

Davor Grgić, Mario Matijević, Paulina Dučkić, Radomir Ječmenica

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

davor.grgic@fer.hr

 

NPP Krsko changed long term spent fuel storage management and decided to store part of the fuel in Spent Fuel Dry Storage (SFDS) installation. NPP Krsko SFDS project is underway. As part of the radiological evaluation of the project we have performed the radiological assessment using hybrid shielding methodology based on MCNP/ADVANTG and SCALE computer codes. The shielding analysis encompassed several Monte Carlo models for quantification of neutron-gamma radiation fields and dose rates: transport cask (HI-TRAC), storage cask (HI-STORM) and dry storage building (DSB). The cask models were developed from reference Holtec International (HI) data and included all of the necessary geometry and material details. Besides such detailed models, important to pin-point important radiological pathways for radiation streaming, simplified “benchmark” models were also developed by HI to verify obtained results and used methodology and to understand impact of design features of storage cask on neutron and gamma dose rates in its vicinity. The benchmark models were used as part of the independent review process to compare methodologies used in the design of the storage cask and the methodologies used in the review, too. In this paper we have presented the results of our calculations related to two benchmark models (both MCNP6 and SCALE calculations) and the calculation of full storage cask model to demonstrate the difference between the simplified and real model. The computer code VisIt (Lawrence Livermore National Laboratory) was also used to provide 3D visualization of deterministic results from the SN code Denovo, which proved to be useful in interpretation of variance reduction parameters for these complex shielding calculations. ORNL meshview utility developed as part of SCALE project was used for visualization of MCNP meshtal files.






12.09.2019 11:30 Radioactive Waste Management

Radioactive Waste Management – 901

Characterizing the Radioactivity of the Concrete Shielding during Decommissioning of the LFR

Perry Young

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

young@nrg.eu

 

The Low Flux Reactor (LFR) at the NRG Petten site has undergone decommissioning in the past years. Gamma spectroscopy measurements were made to ascertain the radioactivity of the LFR concrete shielding. This is an opportunity to compare these measurements against activation calculation results. Being able to predict via calculation the extent to where the concrete is radioactive permits better planning of the decommissioning and estimation of its costs.

MCNP6 [1] is used with a detailed 3D model [2] of the LFR to calculate the flux and spectrum in the concrete shield core extractions. The flux in the samples is then passed onto FISPACT-II [3] for activation calculation. Variance reductions techniques, particularly the Weight Window Mesh, are used to propagate the neutrons in this deep penetration shielding problem; 3 metres deep concrete.

For the calculation to be accurate, it was necessary to obtain measurements of the chemical compositions of the concrete. This was done for several free release core extraction samples. These compositions are needed for the transport calculation in the concrete (MCNP) and the activation calculation (FISPACT). Primary elements (Ba, Fe, O) are determined by XRF and trace elements (Mn, Co, Eu) by ICP.

Nominal C/M results will be presented along with an analysis of the impact of perturbing certain key parameters that have high uncertainties, notably Hydrogen content of the concrete.

REFERENCES
[1] T. Goorley, MCNP6.1.1-Beta Release Notes, LA-UR-14-24680, 2014
[2] W.E. Freudenreich, “Detailed LFR Core Model for MCNP”, ECN-I-94-057, ECN-Petten, July 1995
[3] J-C C. Sublet, et al, The FISPACT-II User Manual, UKAEA-R(11)11 Issue 7, UK Atomic Energy Authority, 2015






12.09.2019 11:50 Radioactive Waste Management

Radioactive Waste Management – 902

Low and Intermediate Level Waste (LILW) repository in Slovenia

Špela Mechora, Sandi Viršek

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

spela.mechora@arao.si

 

In the paper the activities on the project of Low and Intermediate Level Waste (LILW) repository and its disposal concept will be presented. The basic purpose of LILW repository is to prevent the migration of radionuclides into the environment. Slovenia started a siting procedure for a LILW repository in 2004. In the process the public has also been intensively involved, the location Vrbina site in the municipality Krško was selected in 2009, accepted and confirmed by the government. In 2009, a preliminary design was prepared for the proposed concept of disposal, followed by several optimisation studies. A near-surface silo concept for the future Slovenian LILW repository was chosen. The licensing phase started. In 2014 the site investigation was finalised and a „Feasibility study“ – formaly approved next investment phase (until cosntruciton permit), was approved by the Competent Ministry. In 2016 the elaboration of a Report on environmental impacts for a LILW repository was ordered. This represents the basis for initiating an administrative procedure for acquiring an environmental consent for the intended activity (EIA phase) and a construction permit. The elaborated and reviewed Report on environmental impacts for a LILW repository was supplemented in 2018. Application for issuing an environmental consent was accepted by the Environmental Agency of the Republic of Slovenia. The Draft of preliminary consent for nuclear and radiation safety was issued in April 2019, which is one step in an administrative procedure for issuing an environmental consent. At the moment, public hearing (national and neighbouring countries) is underway.






12.09.2019 12:10 Radioactive Waste Management

Radioactive Waste Management – 903

Reprocessing, a sustainable used fuel management strategy

Assia Talbi1, Peter Breitenstein2

1ORANO, 1 Place Jean Millier, 92400 Courbevoie, France

2AREVA, 1 place Jean Millier, 92084 Paris La Defense, France

assia.talbi@orano.group

 

Used nuclear fuels generated by the operation of Nuclear Power Plants (NPP) need to be managed in a safe, responsible and effective way.
Whereas utilities managing several NPP can implement large scale used fuel management operations, a single reactor utility will chose solutions adapted for relatively low amount of used fuel.
Currently, there are two different approaches for managing used fuel:
• Open fuel cycle, or “once-through” strategy, where used fuel is considered to be waste and disposed of after wet or dry interim storage;
• Closed fuel cycle, or “reprocessing” strategy, where used fuel is considered as valuable material as it mainly contains reusable uranium and plutonium; such strategy can be implemented directly after in-reactor use or can be put in place after interim storage. Used fuel reprocessing allows the recovery and recycling of 96% of the nuclear material in Mixed OXide (MOX) fuel and Enriched Reprocessed Uranium (ERU) fuel; the remaining 4% of non-recyclable material, as well as cladding and structural elements of fuel assemblies, being conditioned for final disposal. Final residues are durable, stable and homogeneous, and allow a significant reduction of the volume and radiotoxicity.
With more than 34,000 tons of used fuel reprocessed since 1976 and 50 years’ experience in the nuclear industry, Orano, former Areva, supports nuclear utilities in implementing sustainable used fuel management strategies.
This paper aims to present the used fuel reprocessing management option for a single reactor utility, with a focus on the management of residues form reprocessing, and Orano associated available services.






09.09.2019 18:30 Nuclear Power Plant Operation and New Reactor Technologies

Nuclear Power Plant Operation and New Reactor Technologies – 1001

Comparison Study of Natural Circulation Cooldown Capability for Various Reactor Types

Gwang Hyeok Seo

Korea Institute of Nuclear Safety , 62 Gwahak-ro, Yuseong-gu, Daejeon, 34142, South Korea

ghseo@kins.re.kr

 

Natural circulation cooldown (NCC) is one of required strategies to operate commercial nuclear power reactors (NPPs) safely with suitable subcooling from normal operation to cold shutdown in the event of transients. In response to the NCC operation, the USNRC requires that only safety-grade equipment or systems can be used when onsite or offsite power is available assuming a single failure. In this regards, a proper evaluation of the NCC capability for a certain NPP is essentially demanded. In South Korea, several kinds of NPPs are currently operated or planned such as OPR1000, APR1400, and SMART, a Small Modular Reactor (SMR) recently developed. In this study, the NCC capability and characteristics of the new SMR are evaluated and the analysis results are compared to other kinds of commercial NPPs.
In order to evaluate the effects of design parameters on the plant, the MARS-KS code, a safety analysis system code, is utilized. Several significant components or systems are carefully selected and modeled to simulate the system behavior under the NCC conditions. For example, Residual Heat Removal System (RHR) or Shutdown Cooling System (SCS) provides primary system cooling through decay heat removal after RHR entry conditions. Atmospheric Dump Valves (ADVs) are used to release heat in the secondary system. Auxiliary feedwater flow is carefully controlled to maintain steam generate (SG) levels. The NCC simulation is typically performed as follow: (1) After the reactor trip, the NSSS is stabilized at hot standby conditions for 4 hrs. (2) The RCS is cooled down and depressurized with a gradual increase in subcooling. (3) Finally, the code calculation is terminated when the RHR entry conditions are attained. After the NCC simulation, a comparison work is carried out with open literature data on the NCC analysis on other kinds of reactors. With the NCC simulation and comparison study, the system behavior and the effect of design parameters of the new SMR is evaluated.






09.09.2019 18:50 Nuclear Power Plant Operation and New Reactor Technologies

Nuclear Power Plant Operation and New Reactor Technologies – 1011

Planning Of Modifications Using Laser Scanning Technology

Primož Hostnik, Rajmund Mlakar

SIPRO Inženiring d.o.o., Cesta krških žrtev 135c, 8270 Krško, Slovenia

primoz.hostnik@sipro-inzeniring.si

 

When planning new systems within existing technological systems, investors, designers and installers face a lack of data on the existing situation. Documentation of the existing situation is often deficient, uncoordinated. Using traditional methods it is necessary to carry out activities as review of archives, field measurements, etc. that are time consuming and the effect is limited before starting the planning. Some areas can be accessed due to high temperature, moisture, dust, radiation, chemicals, etc. only when system is out of operation. Due to lack of information there is great risk of errors in design, collisions with existing structures, which must be solved during installation. This can consequently result additional costs, delays and safety issues.

The use of laser scanning technology makes it possible to quickly capture large amounts of data on the existing state with great accuracy. The information is transferred to a digital model that is easily accessible to all participants in the design. All participants can share information via common digital model, which allow progress control, interferences detection and consequently reduces risk of unplanned events during installation.

After finished installation, digital model can be upgraded with non-geometrical data, which is useful for operation and maintenance of the system.

Target group of customers are investors in energetics and industry that are focused in modifications, upgrades and prolonging life time of existing plants. In last decade investments in new plants was reduced due to unstable economic conditions in energetics and process industry. On the other hand, portion of brown-field projects (modifications, upgrades) is increasing.

CASE STUDY
Company Sipro Inženiring carried out following services in preparation phase for installation of modification 1029-RH-L at NEK for contractor IDOM:
– 3D laser scanning of affected rooms.
– Processing of data in form of »Point Clouds« and implementation with 3D model.
– Clash detection between existing structures and new design.
– Simulation of transport of main equipment, as main pump, heat exchanger, MOV.
Results and key benefits are:
– Optimizing work process in terms of human resources, installation timeline.
– Recognizing critical points of installation (clashes, transport, lifts).
– Reducing risks during installation.
– Reducing costs of installation.






09.09.2019 19:10 Nuclear Power Plant Operation and New Reactor Technologies

Nuclear Power Plant Operation and New Reactor Technologies – 1003

Deployment of Small Modular Reactors

Helena Janžekovič, Andreja Peršič

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

helena.janzekovic@gov.si

 

Small Modular Reactors (SMRs) have been developed for decades in parallel with the development of large nuclear reactors. Today there are about 50 different designs and concepts of SMRs globally. As reported by the IAEA most of them are in various developmental stages and some are claimed as being near-term deployable. It seems that security issues which should be taken into account when countries planning their energy policy as well as climate changes gave additional motives to consider introducing SMRs as a part of the strategic nuclear energy programmes. In addition, after the Fukushima accident in 2011 the environmental impact associated with severe accidents at a site of large nuclear power plants gave an addition boost to the development of SMRs.
Today several countries, including EU countries plan to initiate or expand their energy programmes with SMRs, or consider including design of SMRs in their long-term programmes. Some regulatory authorities already received first applications for licence of such facilities, e.g. the very first Canadian SMR licence has been submitted in April 2019. The article analyses a list of the advantages of SMR which includes siting flexibility for locations unable to accommodate traditional larger reactors, possibility to combine nuclear with alternative energy sources, environmental impact and enhanced non-proliferation just to mention some. It tackles also disadvantages associated with SMRs.
The article gives an overview of international activities, e.g. EC Horizon 2020 work programme for 2019-2020 includes topic Support for safety research of Small Modular Reactors, IAEA SMR Regulator’s Forum has been established as well as IAEA Project RER/2/014 – Facilitating Capacity Building for Small Modular Reactors: Technology Developments initiated in 2018. The research project Towards European Licensing of Small Modular Reactors – ELSMOR led by VTT, Finland, has been launched. A brief overview of national activities is given e.g. in UK, Russia and USA. In particular, the article gives overview of regulatory frameworks developed to serve licensing SMRs in countries already prepared for introducing SMRs in very near future. An overview of a current situation, i.e. in 2019, is given as well as foreseen development in next years. The needed regulatory framework to be applicable for countries with relatively small nuclear programme is discussed.






10.09.2019 17:10 Regulatory Issues, Sustainability and Education

Regulatory Issues, Sustainability and Education – 1101

Conducting Trainings of SNSA Emergency Response Team

Igor Sirc, Metka Tomažič

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

metka.tomazic@gov.si

 

Keywords: nuclear safety, emergency preparedness, emergency response team, training, exercise, adult learning

Abstract

Slovenian Nuclear Safety Administration’s (SNSA) main role is ensuring nuclear and radiological safety in the country. In case of an emergency, its key role is to assess the situation and to provide advice and give recommendations to the Civil Protection Commander of the Republic of Slovenia. Its further roles are information of the public in the first phases of an emergency, information and communication with authorities abroad and international organizations, including ensuring the implementation of international assistance procedures, etc. For effective implementation of designated tasks, during an emergency the SNSA sets up the Emergency Response Team. Since the tasks of the SNSA staff during the emergency differ a lot from their everyday tasks, regular and quality trainings are very important. To be prepared as much as possible, many trainings of different types are conducted.
Trainings are in principle divided into initial (basic) trainings and periodic refreshments, they differ also regarding the target audience (for all members of the SNSA Emergency Response Team or for limited groups) and regarding subject matters (all tasks for a specific position or specific issue for a specific position) and are carried out as lectures, on-the job trainings and as participation in different types of exercises.
All trainings and exercises at the SNSA are planned a year in advance by the Emergency Preparedness Division in collaboration with Monitoring Section and Nuclear Safety Division. Once the plan is approved by the Management of the SNSA, it is uploaded into “infoNUID”- the information system, developed specifically to manage the emergency response competences of the SNSA. The Emergency Preparedness Division is responsible for implementation of this plan by organizing the participation of the Emergency Response Team in exercises and trainings, conducting specific trainings, analyzing the conducted exercises and most of all by identifying and monitoring the implementation of improvements needed based on lessons learned either from exercises or trainings.
Implementation of trainings and exercises as well as efficient maintenance of the emergency procedures and care for the dedicated equipment and tools are essential part of the never-ending emergency preparedness process. Considering the nature of nuclear emergencies, especially their rarity on one side and possible devastating consequences on the other, no one can ever claim that the absolute preparedness has been achieved. Continuous improvement is therefore essential and better preparedness of the SNSA can be achieved also through regular and successful trainings of its Emergency Response Team.






10.09.2019 17:50 Regulatory Issues, Sustainability and Education

Regulatory Issues, Sustainability and Education – 1103

National Strategy for Research and Development in the Area of a Use of Nuclear Energy and Radiation Sources in Slovenia

Barbara Vokal Nemec, Andreja Peršič, Helena Janžekovič

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

barbara.vokal-nemec@gov.si

 

Before second world war research related to nuclear and radiation was flourishing in only few countries, while after this war many countries established so-called “nuclear institute” with clear strategic agenda related research in military use of nuclear industry. In last seventy years many of such institutes went through substantial transformation and are nowadays dedicated to research in non-military use of nuclear industry. In some cases researchers have been reoriented into areas far away from any application of a use of ionizing radiation. As after the Fukushima accident phase out of nuclear energy has been announced in some country and this political decision influences also research programmes as they are now more oriented on decommissioning than one the development of new type of reactors.
A use of radiation sources does not show any decline, moreover, a use of radiation sources in many field is an inevitable part of modern life such as treatment of oncology patients with ionizing radiation or treatment of industrial products with accelerators to change their properties. In this context the country should revise their programmes related to research needed. In the EU Euratom Directives tackles research, e.g. regarding in Article 8c of the Council Directive 2014/87/EURATOM so-called “ most recent research results” should be taken into account by the licence holder when conducting initial assessment and periodic safety reviews while Requirement 1 of the IAEA GSR Part 1 (Rev 1) require that the national policy and strategy for safety tackles includes provision and framework for research and development among others.
The strategy for research and development in the area of a use of nuclear energy and radiation sources is a country specific document based on the overall present and future framework. In Slovenia the very first national strategy called SNSA Strategy for Research and Development was prepared in 2013. The article describes challenges related to the updating of this strategy which started in 2018. A small team of expert has been nominated to do this task, questionnaire dedicated to analyse present situation in the country has been prepared and a one day meeting has been organised with all stakeholders presenting their past, present as future research programmes. The article aims helping other regulatory authorities managing relatively small nuclear programme to handle such a challenging task.